Ueno, Yasuhiro*; Aoki, Masaharu*; Fukao, Yoshinori*; Higashi, Yoshitaka*; Higuchi, Takashi*; Iinuma, Hiromi*; Ikedo, Yutaka*; Ishida, Keiichi*; Ito, Takashi; Iwasaki, Masahiko*; et al.
Hyperfine Interactions, 238(1), p.14_1 - 14_6, 2017/11
Strasser, P.*; Aoki, Masaharu*; Fukao, Yoshinori*; Higashi, Yoshitaka*; Higuchi, Takashi*; Iinuma, Hiromi*; Ikedo, Yutaka*; Ishida, Keiichi*; Ito, Takashi; Iwasaki, Masahiko*; et al.
Hyperfine Interactions, 237(1), p.124_1 - 124_9, 2016/12
Kokubo, Nobuhito*; Miyahara, Hajime*; Okayasu, Satoru; Nojima, Tsutomu*
Journal of the Physical Society of Japan, 84(4), p.043704_1 - 043704_4, 2015/04
We report on the direct observation of vortex states confined in equilateral and isosceles triangular dots of weak pinning amorphous superconducting thin films with a scanning superconducting quantum interference device microscope. The observed images illustrate not only pieces of a triangular vortex lattice as commensurate vortex states, but also incommensurate vortex states including metastable ones. We comparatively analyze vortex configurations found in different sample geometries and discuss the symmetry and stability of commensurate and incommensurate vortex configurations against deformations of the sample shape.
Kokubo, Nobuhito*; Okayasu, Satoru; Nojima, Tsutomu*; Tamochi, Hirotaka*; Shinozaki, Bunju*
Journal of the Physical Society of Japan, 83(8), p.083704_1 - 083704_5, 2014/08
Ohgama, Kazuya; Oki, Shigeo; Sugino, Kazuteru; Okubo, Tsutomu
Journal of Nuclear Science and Technology, 51(4), p.558 - 567, 2014/04
Core characteristics of a sodium-cooled fast breeder reactor (FBR) with 750 MWe output using highly decontaminated uranium and plutonium and highly minor-actinide-containing compositions were evaluated using the fast reactor cross-section set generated by the new Japanese nuclear data library JENDL-4.0. The core characteristics were compared with those obtained using the unified cross-section set ADJ2000R in order to investigate the differences between both the results. The effects on the core characteristics caused by the differences in the nuclear data of important reactions and nuclides in the cross-section sets were analyzed by a burnup sensitivity analysis. It was confirmed that adopting JENDL-4.0 to the FBR core design improves the breeding ratio, the burnup reactivity, and the reactivity control balance, because of the differences in the capture cross-sections of U-238 and Pu-239 of both the libraries. The difference in the sodium void reactivity evaluated with both the libraries was less than 1% because the increase caused by the differences in the elastic scattering cross-sections of sodium, the inelastic scattering cross-section, and the -average value of U-238 was practically cancelled out by the decrease caused by the differences in the capture cross-sections of Pu-239, the inelastic scattering cross-section of iron, and the capture cross- sections of Am-241.
Sasaki, Kenji*; Naito, Katsuaki*; Oki, Shigeo; Okubo, Tsutomu; Kotake, Shoji*
Progress in Nuclear Science and Technology (Internet), 4, p.94 - 98, 2014/04
This paper evaluates the amount of activation of the secondary sodium in Direct Heat Exchanger (DHX) by neutrons leaked from the core, the radioactivity density, and the dose rate around the secondary sodium pipes in Direct Reactor Auxiliary Cooling System (DRACS) and confirms that the requirements in radioactivity free areas are satisfied by improving the exactness of calculation model with Monte Carlo Methodology.
Kawashima, Katsuyuki; Sugino, Kazuteru; Oki, Shigeo; Okubo, Tsutomu
Nuclear Technology, 185(3), p.270 - 280, 2014/03
Although the sodium void reactivity is limited up to 6 dollars in the current JSFR design, it should be significant to perform design studies of the low sodium void reactivity core besides the reference design, to increase the design margin considering any influence of the TRU fuel compositions. In this study, the BUMPY core is proposed as the low sodium void core concept, in which the partial-length fuels with upper sodium plenum are interspersed within the core, causing the steps in fuel length in the neighboring fuel assemblies. The void reactivity is considerably reduced due to the upward and lateral neutron leakage from the fuel region to the sodium plenum upon voiding. The BUMPY core is applied to the JSFR design. The calculated void reactivity of the BUMPY core is 2.5 dollars, which is considerably reduced from 5.3 dollars for the reference core. Moreover, the Doppler coefficient is almost the same as the reference core.
Meiliza, Y.; Oki, Shigeo; Kawashima, Katsuyuki; Okubo, Tsutomu
Progress in Nuclear Energy, 70, p.270 - 278, 2014/01
Meiliza, Y.; Oki, Shigeo; Kawashima, Katsuyuki; Okubo, Tsutomu
Journal of Nuclear Science and Technology, 50(6), p.615 - 628, 2013/05
Aoto, Kazumi; Chikazawa, Yoshitaka; Okubo, Tsutomu; Okada, Keizo*; Ito, Takaya*
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Safe Technologies and Sustainable Scenarios (FR-13) (USB Flash Drive), 10 Pages, 2013/03
Overview of Japan Sodium-cooled Fast Reactor (JSFR) development status and reflection of lessons learned from the TEPCO's Fukushima Dai-ichi Nuclear Power Plant (1F) accident have been summarized. JSFR was recognized as a promising next generation nuclear reactor. Even though the JSFR safety design already took into account measures against severe accident situations and passive safety features such as passive shutdown system and natural convection decay heat removal systems in the 2010 design version, it is become aware of importance of design measures against severe accidents and extreme external events by the 1F accident. As recent activities, external hazard evaluations and design improvements reflecting lessons learned from 1F accident have been conducted. This paper also discusses importance of development of global safety design criteria and international Research and Development cooperation on safety design measures.
Hayafune, Hiroki; Kato, Atsushi; Chikazawa, Yoshitaka; Okubo, Tsutomu; Sagawa, Hiroshi*; Shimakawa, Yoshio*
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Safe Technologies and Sustainable Scenarios (FR-13) (USB Flash Drive), 11 Pages, 2013/03
Evaluation of earthquake and tsunami on JSFR has been analyzed. For seismic design, safety components are confirmed to maintain their functions even against recent strong earthquakes. As for Tsunami, some parts of reactor building might be submerged including component cooling water system whose final heat sink is sea water. However, in the JSFR design, safety grade components are independent from component cooling water system (CCWS). The JSFR emergency power supply adopts a gas turbine system with air cooling, since JSFR does not basically require quick start-up of the emergency power supply thanks to the natural convection DHRS. Even in case of long station blackout, the DHRS could be activated by emergency batteries or manually and be operated continuously by natural convection.
Uematsu, Mari Mariannu; Sugino, Kazuteru; Kawashima, Katsuyuki; Okano, Yasushi; Yamaji, Akifumi; Naganuma, Masayuki; Oki, Shigeo; Okubo, Tsutomu; Ota, Hirokazu*; Ogata, Takanari*; et al.
JAEA-Research 2012-041, 126 Pages, 2013/02
The characteristics of sodium-cooled metal fuel core compared to MOX fuel core are given by its higher heavy metal density and superior neutron economy. By taking advantage of these characteristics and allowing flexibility in metal fuel specification and core design conditions as sodium void reactivity and bundle pressure drop, core design with high burnup, high breeding ratio and low fuel inventory features will be achievable. On ground of the major achievements in metal fuels utilization as driver fuels in sodium fast reactors in U.S., the metal fuel core concept is selected as a possible alternative of MOX fuel core concept in FaCT project. This report describes the following items as a result of the joint study on "Reactor core and fuel design of metal fuel core of sodium-cooled fast reactor" conducted by JAEA and CRIEPI during 4 years from fiscal year 2009 to 2012.
Oki, Shigeo; Naganuma, Masayuki; Okubo, Tsutomu; Tanaka, Kenya
Journal of Nuclear Science and Technology, 50(1), p.59 - 71, 2013/01
Yamaji, Akifumi; Nakano, Yoshihiro; Uchikawa, Sadao; Okubo, Tsutomu
Nuclear Technology, 179(3), p.309 - 322, 2012/09
HC-FLWR effectively utilizes the uranium (U) and the plutonium (Pu) resources by achieving a fissile Pu conversion ratio of 0.84 without a significant technical gap from the current BWR technology. In this study, a new core design concept for HC-FLWR has been developed to achieve the conversion ratio of 0.95. The concept of the FLWR/MIX fuel assembly, which had been originally proposed for tight fuel bundle, was used to raise the conversion ratio without deteriorating the core void reactivity characteristics. For a semi-tight fuel rod lattice with rod clearance of 0.20 to 0.25 cm, the design ranges of the conversion ratio and the average discharge burnup are 0.91 to 0.94 and 53 to 49 GWd/t, respectively. The conversion ratio can be raised to 0.97 by increasing the U enrichment from 4.9 to 6.0 wt%. Two representative core designs and one alternative design option have been obtained. Hence, the flexibility of HC-FLWR concept to achieve the conversion ratio of 0.84 to 0.95 has been revealed.
Ohgama, Kazuya; Oki, Shigeo; Okubo, Tsutomu
JAEA-Conf 2012-001, p.21 - 26, 2012/07
Maruyama, Shuhei; Oki, Shigeo; Okubo, Tsutomu; Kawashima, Katsuyuki; Mizuno, Tomoyasu
Journal of Nuclear Science and Technology, 49(6), p.640 - 654, 2012/06
This study shows the good correlations in FBR core characteristics, and find out the mechanism of their correlations with the aid of sensitivity analyses. It has been clarified that Doppler coefficient turns to have the correlations with the other core characteristics by considering the constraint of the criticality requirement for fuel composition variations. The finding of the correlations makes easy to specify the ranges of core reactivity control and core safety properties which are important for core design in determining core specification and performance. It gives significant information for FBR core design in the transition stage. Moreover, as an application of the above-mentioned correlations, a simplified burnup reactivity index is developed for rapid and rational estimation of the core characteristic variations. By using this index and the correlations, the core characteristic variations can be estimated for various fuel compositions without repeating core calculations.
Uchikawa, Sadao; Okubo, Tsutomu; Nakano, Yoshihiro
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 8 Pages, 2012/00
An advanced LWR with hard neutron spectrum named FLWR is a BWR-type reactor with a core consisting of hexagonal-shaped fuel assemblies with a triangular tight-lattice fuel rod configuration. It has been proposed in order to ensure sustainable energy supply in the future based on the well-experienced LWR technologies. The reactor concept of the FLWR is designed to utilize the most of the existing Advanced Boiling Water Reactor (ABWR) plant system. Therefore, only the core concept is new. The FLWR aims at effective and flexible utilization of uranium and plutonium resources by adopting a two-stage concept of core designs. The core in the first stage of FLWR is for intensive utilization and conservation of plutonium with no degradation of the isotopic quality of plutonium based on the experience of the current LWR-MOX utilizations. The one in the second stage realizes sustainable multiple plutonium recycling with a high conversion ratio over 1.0. When the technologies and infrastructures for multiple recycling with MOX spent fuel reprocessing are established, the core of the first stage proceeds to the second stage by only changing the fuel assembly design in the same reactor system. The present paper summarizes the recent core design studies of FLWR.
Kawashima, Katsuyuki; Ogawa, Takashi; Oki, Shigeo; Okubo, Tsutomu; Mizuno, Tomoyasu
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 12 Pages, 2012/00
Sodium-cooled fast reactor core design considerations are made to improve the proliferation resistance by focusing on the plutonium generated in the UO blanket in the frame of the Fast Reactor Cycle Technology Development (FaCT) project. The appropriate design and treatments of the UO blanket help to reduce the intrinsic proliferation potentials. Based on the 1500 MWe FaCT reference core, the three different cores (radial blanket-free core, the core with the low-enriched MOX fuel, and the core with MA-doped UO fuel) are configured to meet the provisional proliferation resistance criteria as well as the core performance targets.
Nakano, Yoshihiro; Okubo, Tsutomu
Annals of Nuclear Energy, 38(12), p.2689 - 2697, 2011/12
The isotopic composition and amount of Pu in spent fuel from high burnup BWR and PWR (HB-BWR, HB-PWR), each with 70 GWd/t discharge burnup and 6% U enrichment were estimated to evaluate FBR fuel composition in the transition period from LWRs to FBRs. The HB-BWR employs spectral shift rods. The fraction of fissile Pu (Puf) in HB-BWR spent fuel after 5 years cooling is 62%, which is larger than that of conventional BWRs with burnup of 45 GWd/t, because of the spectral shift operation. The amount of Pu produced in the HB-BWR is also larger than that produced in a conventional BWR. The HB-PWR uses a wider pitch 1717 assembly to optimize neutron slowing down. The Puf fraction of HB-PWR spent fuel after 5 years cooling is 56%, which is smaller than that of conventional PWRs with burnup of 49 GWd/t, mainly because of the wider pitch. The amount of Pu produced in the HB-PWR is also smaller than that in conventional PWRs.
Uchikawa, Sadao; Okubo, Tsutomu; Nakano, Yoshihiro
Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 8 Pages, 2011/12
An advanced LWR with hard neutron spectrum, named FLWR, aims at efficient and flexible utilization of nuclear resources by evolving its fuel assembly design under the same core configuration, mainly corresponding to available fuel cycle technologies and related infrastructures. The paper summarizes an evolution process of the FLWR fuel assembly design toward a sustainable fuel cycle by dividing the reactor operation into three stages, that is, the one based on the current LWR MOX fuel cycle infrastructure such as reprocessing of UO spent fuel and fabrication of MOX fuel, the one for transitioning from the LWR fuel cycle to the FR fuel cycle, and the one based on the FR fuel cycle infrastructures such as MOX spent fuel reprocessing.