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Journal Articles

Development of fuel performance analysis code, BISON for MOX, named Okami; Analyses of pore migration behavior to affect the MA-bearing MOX fuel restructuring

Ozawa, Takayuki; Hiroka, Shun; Kato, Masato; Novascone, S.*; Medvedev, P.*

Journal of Nuclear Materials, 553, p.153038_1 - 153038_16, 2021/09

 Times Cited Count:0 Percentile:0.01(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.

JAEA Reports

Estimation of exchange time for neutron startup sources of HTTR

Ono, Masato; Kozawa, Takayuki; Fujimoto, Nozomu*

JAEA-Technology 2019-012, 15 Pages, 2019/09


The High Temperature Engineering Test Reactor has a neutron source of $$^{252}$$Cf to start up the reactor and to confirm count rates of wide range monitors. The half-life of $$^{252}$$Cf is short, about 2.6 years, so it is necessary to replace at an appropriate time. In order to estimate the period to replace, it is necessary to consider not only the half-life but also the fluctuation of the count rate of the wide range monitor to prevent alarm. For that reason, the method has been derived to predict a minimum count rate from relationship between the count rate and the standard deviation of the count rate of the wide range monitors. As a result of predicting the count rate using this method, it was found that the minimum count rate reaches to 3.0cps in 2022 and 1.5 cps in 2024. Therefore, it is necessary to exchange $$^{252}$$Cf by 2024.

Journal Articles

Analysis of fast reactor fuel irradiation behavior in the MA recycle system

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 $$sim$$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.

Journal Articles

Current status of the next generation fast reactor core & fuel design and related R&Ds in Japan

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.

Journal Articles

Fuel restructuring behavior analysis of MA-bearing MOX fuels irradiated in a fast reactor

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 $$sim$$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.

JAEA Reports

A Study on laser welding of ferritic/martensitic steel (PNC-FMS) for fast reactor fuel assemblies

Kono, Fumiaki; Sogame, Motomu; Yamada, Tomonori; Shobu, Takahisa; Naganuma, Masayuki; Ozawa, Takayuki; Muramatsu, Toshiharu

JAEA-Technology 2015-004, 57 Pages, 2015/03


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 $$^{circ}$$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.

JAEA Reports

Development of annular fuel design code CEPTAR-D; Study on applicability for PCMI stress evaluation

Kamei, Miho; Ozawa, Takayuki

JAEA-Technology 2014-033, 36 Pages, 2014/11


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.

Journal Articles

Establishment of control technology of the HTTR and future test plan

Honda, Yuki; Saito, Kenji; Tochio, Daisuke; Aono, Tetsuya; Hirato, Yoji; Kozawa, Takayuki; Nakagawa, Shigeaki

Journal of Nuclear Science and Technology, 51(11-12), p.1387 - 1397, 2014/11

 Times Cited Count:1 Percentile:10.7(Nuclear Science & Technology)

The operational experiments of the HTTR would be useful for future high-temperature gas-cooled reactors (HTGRs). Main PID control constants of the HTTR are selected with reasonably damped characteristics and without undershoot or overshoot. For utilization the HTGR as a commercial reactor, it should be demonstrated that the HTGR system can supply stable heat to a heat utilization system for the long-term operation. The control characteristics in the long-term high-temperature operation are evaluated by the result of operation performed in 2010. In addition, from a viewpoint of HTGRs with heat utilization system, a future possibility of the experiments for heat utilization design is examined.

Journal Articles

Development and verification of the thermal behavior analysis code for MA containing MOX fuels

Ikusawa, Yoshihisa; Ozawa, Takayuki; Hiroka, 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.

Journal Articles

Applicability of iron phosphate glass medium for loading NaCl originated from seawater used for cooling the stricken power reactors

Kobayashi, Hidekazu; Amamoto, Ippei; Yokozawa, Takuma; Yamashita, Teruo; Nagai, Takayuki; Kitamura, Naoto*; Takebe, Hiromichi*; Mitamura, Naoki*; Tsuzuki, Tatsuya*

Proceedings of 15th International Conference on Environmental Remediation and Radioactive Waste Management (ICEM 2013) (CD-ROM), 6 Pages, 2013/09

no abstracts in English

Journal Articles

Behaviour of IPG waste forms bearing BaSO$$_{4}$$ as the dominant sludge constituent generated from the treatment of water used for cooling the stricken power reactors

Amamoto, Ippei; Kobayashi, Hidekazu; Yokozawa, Takuma; Yamashita, Teruo; Nagai, Takayuki; Kitamura, Naoto*; Takebe, Hiromichi*; Mitamura, Naoki*; Tsuzuki, Tatsuya*

Proceedings of 15th International Conference on Environmental Remediation and Radioactive Waste Management (ICEM 2013) (CD-ROM), 8 Pages, 2013/09

The great amount of water used for cooling the stricken power reactors at Fukushima Dai-ichi has resulted in accumulation of "remaining water". As the remaining water is subsequently contaminated by FPs, etc., it is necessary to decontaminate it in order to reduce the volume of liquid radioactive waste and to reuse it again for cooling the reactors. Various techniques are being applied to remove FP, etc. and to make stable waste forms. One of the methods using the iron phosphate glass as a medium is being developed to stabilize the strontium-bearing sludge whose main component is BaSO$$_{4}$$. From the results hitherto, the iron phosphate glass is regarded as a potential medium for the target sludge.

Journal Articles

Decontamination of outdoor school swimming pools in Fukushima after the nuclear accident in March 2011

Saegusa, Jun; Kurikami, Hiroshi; Yasuda, Ryo; Kurihara, Kazuo; Arai, Shigeki; Kuroki, Ryota; Matsuhashi, Shimpei; Ozawa, Takashi; Goto, Hiroaki; Takano, Takao; et al.

Health Physics, 104(3), p.243 - 250, 2013/03

 Times Cited Count:3 Percentile:29.69(Environmental Sciences)

After the Nuclear accident on March 2011, water discharge from many outdoor swimming pools in the Fukushima prefecture was suspended out of concern that radiocesium in the pool water would flow into farmlands. We have reviewed the existing flocculation method for decontaminating pool water and established a practical decontamination method by demonstrating the process at several pools in the Fukushima prefecture.

Journal Articles

Development of small specimen test techniques for the IFMIF test cell

Wakai, Eiichi; Kim, B. J.; Nozawa, Takashi; Kikuchi, Takayuki; Hirano, Michiko*; Kimura, Akihiko*; Kasada, Ryuta*; Yokomine, Takehiko*; Yoshida, Takahide*; Nogami, Shuhei*; et al.

Proceedings of 24th IAEA Fusion Energy Conference (FEC 2012) (CD-ROM), 6 Pages, 2013/03

Journal Articles

New result in the production and decay of an isotope, $$^{278}$$113 of the 113th element

Morita, Kosuke*; Morimoto, Koji*; Kaji, Daiya*; Haba, Hiromitsu*; Ozeki, Kazutaka*; Kudo, Yuki*; Sumita, Takayuki*; Wakabayashi, Yasuo*; Yoneda, Akira*; Tanaka, Kengo*; et al.

Journal of the Physical Society of Japan, 81(10), p.103201_1 - 103201_4, 2012/10

 Times Cited Count:142 Percentile:97.35(Physics, Multidisciplinary)

An isotope of the 113th element, $$^{278}$$113, was produced in a nuclear reaction with a $$^{70}$$Zn beam on a $$^{209}$$Bi target. We observed six consecutive $$alpha$$ decays following the implantation of a heavy particle in nearly the same position in the semiconductor detector, in extremely low background condition. The fifth and sixth decays are fully consistent with the sequential decays of $$^{262}$$Db and $$^{258}$$Lr both in decay energies and decay times. This indicates that the present decay chain consisted of $$^{278}$$113, $$^{274}$$Rg (Z = 111), $$^{270}$$Mt (Z = 109), $$^{266}$$Bh (Z = 107), $$^{262}$$Db (Z = 105), and $$^{258}$$Lr (Z = 103) with firm connections. This result, together with previously reported results from 2004 and 2007, conclusively leads the unambiguous production and identification of the isotope $$^{278}$$113, of the 113th element.

Journal Articles

Production and decay properties of $$^{264}$$Hs and $$^{265}$$Hs

Sato, Nozomi; Haba, Hiromitsu*; Ichikawa, Takatoshi*; Kaji, Daiya*; Kudo, Yuki*; Morimoto, Koji*; Morita, Kosuke*; Ozeki, Kazutaka*; Sumita, Takayuki*; Yoneda, Akira*; et al.

Journal of the Physical Society of Japan, 80(9), p.094201_1 - 094201_7, 2011/09

 Times Cited Count:11 Percentile:61.11(Physics, Multidisciplinary)

Decay properties of $$^{264}$$Hs and $$^{265}$$Hs produced in the $$^{207,208}$$Pb($$^{58}$$Fe, $$xn$$) [$$x$$=1, 2] reactions were studied using a gas-filled recoil ion separator at the linear accelerator facility of RIKEN. A total of 6 decay chains were assigned to $$^{264}$$Hs. Cross sections for the $$^{264}$$Hs production in the $$^{208}$$Pb($$^{58}$$Fe,$$2n$$) and $$^{207}$$Pb($$^{58}$$Fe,$$n$$) reactions were measured to be $$2.8^{+6.5}_{-2.3}$$pb and $$6.9^{+4.4}_{-3.1}$$pb, respectively. The isotope $$^{264}$$Hs decayed with a half-life of $$0.751^{+0.518}_{-0.218}$$ms by $$alpha$$-particle emission and spontaneous fission. The $$alpha$$-particle energy of $$^{264}$$Hs was observed at 10.61$$pm$$0.04 and 10.80$$pm$$0.08 MeV. The spontaneous fission branch of $$^{264}$$Hs was found to be $$17^{+38}_{-14}%$$.

Journal Articles

Physical properties and irradiation behavior analysis of Np- and Am-bearing MOX fuels

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 ($$O$$/$$M$$) 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 $$O$$/$$M$$ 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 $$O$$/$$M$$=1.96 was estimated to reach 2680K which was 130K higher than that of $$O$$/$$M$$=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.

Journal Articles

Water transport in polymer electrolyte membranes investigated by dissipative particle dynamics simulation

Sawada, Shinichi; Yamaki, Tetsuya; Ozawa, Taku*; Suzuki, Akihiro*; Terai, Takayuki*; Maekawa, Yasunari

ECS Transactions, 33(1), p.1067 - 1078, 2010/10

 Times Cited Count:6 Percentile:89.92

In order to investigate water transport in polymer electrolyte membranes (PEMs), we calculated the self-diffusion coefficient of water, D$$_{W}$$, by dissipative particle dynamics (DPD) simulation. The simulation target materials are Nafion and radiation-grafted PEMs in the fully-hydrated states. D$$_{W}$$ was obtained by the following steps: (1) molecular modeling with the coarse-grained particles representing groups of several atoms; (2) calculation of the water particle diffusivity, D$$_{W}$$$$^{Particle}$$, in the PEMs; (3) determination of the unit time in the DPD simulation; (4) conversion of D$$_{W}$$$$^{Particle}$$ of the PEMs into D$$_{W}$$ in the standard SI unit. Interestingly, D$$_{W}$$ was found to decrease with the diffusion time period, $$Delta$$t, probably owing to the geometrical confinement effect by water-transport hydrophilic regions. Quantitative analysis of this D$$_{W}$$-$$Delta$$t relationship provided us with the size of hydrophilic regions.

Journal Articles

MOX fuel performance and database development for MOX fuel use in LWRs

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.

Journal Articles

Burn-up effect on MOX fuel thermal conductivity

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.

JAEA Reports

Development of mechanical seal structure for reuse of re-instrumentation device used in fuel irradiation test

Inoue, Shuichi; Yamaura, Takayuki; Saito, Takashi; Ishikawa, Kazuyoshi; Kikuchi, Taiji; Sozawa, Shizuo; Tsuchiya, Kunihiko

JAEA-Technology 2009-076, 33 Pages, 2010/03


In Japan Material Testing Reactor (JMTR), a lot of experiments of fuel irradiation with the power ramping tests have been performed by using the shroud irradiation facility and the Boiling Water Capsule (BOCA). The fuel samples used in these tests were welded to re-instrumentation devices such as thermocouples and FP gas pressures. In this development, the mechanical connection method as "mechanical seal structure", that enables the re-use of re-instrumentation devices, was adopted in order to improve the utilization efficiency of the device. The test samples with mechanical seal structure were fabricated and the confirmatory tests such as He leakage test, thermal cycle test, autoclave test, etc. were carried out. The test samples with the mechanical seal structure showed an excellent result in various confirmatory tests, and the prospect are bright for the re-use of re-instrumentation devices with the mechanical seal structure.

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