Murakami, Tatsutoshi; Kato, Masato; Suzuki, Kiichi; Uno, Hiroki*
Proceedings of 2010 International Congress on Advances in Nuclear Power Plants (ICAPP '10) (CD-ROM), p.1859 - 1865, 2010/06
The sintering behavior of the de-waxed pellets containing about 3000 ppm of carbon was analyzed during the sintering process using thermal gravimetry and dilatometer measurements as a parameter of the ratio of hydrogen partial pressure-to-moisture partial pressure (H/HO) in the sintering atmosphere. The attained O/M ratio and the shrinkage rate increased with decreasing H/HO ratio in the sintering atmosphere. As a result, it is considered that a carbothermic reduction caused the significant decrease of the O/M ratio in the case of the sintering in the atmosphere of high H/HO ratio. In contrast, decrement of O/M ratio could be inhibited by keeping the oxygen potential of the atmosphere high in the case of the sintering in the atmosphere of lower H/HO ratio.
Takeuchi, Kentaro; Kato, Masato; Sunaoshi, Takeo*
Proceedings of 2010 International Congress on Advances in Nuclear Power Plants (ICAPP '10) (CD-ROM), p.1866 - 1873, 2010/06
The effect of the O/M ratio on the sintering behavior of MOX pellets was evaluated from the measurement results of O/M ratio and shrinkage ratio of pellets during the sintering process. Change of O/M ratio was caused by approaching the equilibrium condition which was decided from oxygen potential depending on the H/HO ratio in the sintering atmosphere. The lower the H/HO ratio, the more shrinkage of pellets proceeded at lower temperatures and higher O/M values. The curves of shrinkage rate were observed to have two peaks at 1000 to 1300C and 1400 to 1600C. The shrinkage in the lower temperature range increased with decreasing the H/HO ratio. These results suggested that different mechanisms dominate the sintering behavior at each temperature range. It was confirmed that the sintering behavior of MOX pellet depends on H/HO ratio and changes significantly with O/M ratio.
Kurisaka, Kenichi; Shimakawa, Yoshio*
Proceedings of 2010 International Congress on Advances in Nuclear Power Plants (ICAPP '10) (CD-ROM), p.1220 - 1227, 2010/06
JAEA participates in activities of the Generation IV International Forum's Risk and Safety Working Group (GIF/RSWG). The GIF/RSWG has developed the Integrated Safety Assessment Methodology (ISAM), which consists of five distinct analytical tools; i.e., Qualitative Safety Features Review, Phenomena Identification and Ranking Table (PIRT), Objective Provision Tree (OPT), Deterministic and Phenomenological Analyses (DPA) and Probabilistic Safety Analysis (PSA). Among them, PIRT, OPT, DPA and PSA were applied preliminarily to the Japanese Sodium-cooled Fast Reactor (JSFR) system. The JSFR system includes major innovative safety features such as Self-Actuated Shutdown System (SASS), passive decay heat removal system. PIRT was applied to examination of the reactor safe shutdown by means of SASS during a loss-of-flow accident with a failure of conventional reactor shutdown system. OPTs were developed to assess the structure of safety architecture of the JSFR in an adequate manner based on the defence-in-depth philosophy. Some provisions explicitly shown in the OPT are characterized with the safety design requirements of decay heat removal function. Sufficiency to those requirements was confirmed by DPA. PSA was conducted with analytical models, which were based on those OPT and DPA results. The PSA served to quantification of the level of safety and to the system design improvement in the JSFR.
Takase, Kazuyuki; Shobu, Takahisa; Tsukimori, Kazuyuki; Muramatsu, Toshiharu
Proceedings of 2010 International Congress on Advances in Nuclear Power Plants (ICAPP '10) (CD-ROM), p.1725 - 1732, 2010/06
At the commercial use stage in sodium-cooled fast breeder reactors, securing maintenance and repair better than an equal to that of present water-cooled reactors is needed. Especially a repair technology that secures the plant integrity for long-term operation period becomes indispensable in the heat exchanger tubes of the steam generator that form the boundary of sodium and water coolants. Then the present study focused on the standardization of welding technology with a laser. An experimental study regarding the welding of a stainless steel plate with the laser using fine metal powders is being performed. Moreover, a numerical study is performed to simulate the welding of the fine metal powder on the stainless steel plate. The fine metal powder is made of iron, and is heated by the laser beam, and then melts exceeding the melting temperature. This paper reports the computational results of the welding phenomenon of some metal powders which changes from solid to liquid and liquid to solid. The results were compared with the experimental results qualitatively. It was concluded that a welding simulation is possible and the present numerical approach will be effective.
Chikazawa, Yoshitaka; Uzawa, Masayuki*; Usui, Shinichi*; Tozawa, Katsuhiro*; Kotake, Shoji
Proceedings of 2010 International Congress on Advances in Nuclear Power Plants (ICAPP '10) (CD-ROM), p.10142_1 - 10142_9, 2010/06
Kurome, Kazuya*; Murakami, Hisatomo*; Tsujita, Yoshihiro*; Futagami, Satoshi; Hayafune, Hiroki
Proceedings of 2010 International Congress on Advances in Nuclear Power Plants (ICAPP '10) (CD-ROM), p.637 - 644, 2010/06
Kato, Atsushi; Hirata, Shingo; Chikazawa, Yoshitaka; Uto, Nariaki; Obata, Hiroyuki*; Kotake, Shoji*; Uzawa, Masayuki*
Proceedings of 2010 International Congress on Advances in Nuclear Power Plants (ICAPP '10) (CD-ROM), p.692 - 699, 2010/06
In the JSFR design, a single rotating plug and an upper inner structure (UIS) with a vertically penetrating slit are proposed, so that the fuel handling machine (FHM) can access any subassembly by horizontal movement of the FHM arm in the slit space. As a result of a full-scale mockup test, excellent performance in normal operation has been shown. In this study, from the viewpoint of achieving reliability of the pantograph type FHM, behavior of the FHM mockup have been investigated under abnormal conditions.
Aizawa, Kosuke; Fujita, Kaoru; Hirata, Shingo; Kasahara, Naoto
Proceedings of 2010 International Congress on Advances in Nuclear Power Plants (ICAPP '10) (CD-ROM), p.645 - 652, 2010/06
A conceptual design study of an advanced large-sized (1500MWe class) sodium-cooled fast reactor (named JSFR) has progressed in the FaCT project in Japan. JSFR adopts a selector-valve mechanism for the failed fuel detection and location (FFDL) system. The drive shaft rotates and moves vertically in order to select the channel. And, the drive shaft is in contact with the selector-valve drum by spring load. Thus, a mechanical wear could occur between the drive shaft and the drum of the selector-valve FFDL system. There is concern about manufacturing capability and endurance of the JSFR selector-valve. To demonstrate manufacturing capability and endurance of the JSFR selector-valve, a mock-up was manufactured and an endurance experiment under high temperature sodium has been conducted.
Tanaka, Masaaki; Ohshima, Hiroyuki; Yamano, Hidemasa; Aizawa, Kosuke; Fujisaki, Tatsuya*
Proceedings of 2010 International Congress on Advances in Nuclear Power Plants (ICAPP '10) (CD-ROM), p.1699 - 1708, 2010/06
Numerical simulations with the Unsteady Reynolds Averaged Navier Stokes equation (U-RANS) approach are examined at high Reynolds number conditions in the 1/3 scale water experiment which is a scale model of the hot-leg piping of JSFR. Applicability of the U-RANS approach to the unsteady flow through the short-elbow is confirmed and the pressure fluctuation generation mechanism in relation to the unsteady large scale eddy motion is clarified. Moreover, preliminary numerical simulation of a full-scale model of the hot-leg piping of JSFR is examined and applicability of the URANS approach to the hot-leg piping of JSFR is confirmed.
Sakai, Takaaki; Kotake, Shoji; Aoto, Kazumi; Ito, Takaya*; Kamishima, Yoshio*; Oshima, Jun*
Proceedings of 2010 International Congress on Advances in Nuclear Power Plants (ICAPP '10) (CD-ROM), p.521 - 530, 2010/06
JAEA is now conducting "Fast Reactor Cycle Technology Development (FaCT)" project for the commercialization before 2050s. A demonstration reactor of Japan Sodium-cooled Fast Reactor (JSFR) is planned to start operation around 2025. In the FaCT project, conceptual design study on the demonstration reactor has been performed since 2007 to determine the referential reactor specifications for the next stage design work from 2011 for the licensing and construction. Plant performance as a demonstration reactor for the 1.5 GWe commercial reactor JSFR is being compared between 750 MWe and 500 MWe plant designs. By using the results of conceptual design study, output power will be determined during year of 2010. In this paper, current status of the conceptual design study will be summarized with related research and developments on plant technologies.
Yanagisawa, Kazuaki; Nagano, Koji*
Proceedings of 2010 International Congress on Advances in Nuclear Power Plants (ICAPP '10) (CD-ROM), p.2140 - 2149, 2010/06
(1) The cost of electricity generated by 54 Japanese light water reactors in 2005 is 16,466 million dollars at the supply end and is 42,682 million dollars at the demand end. (2) The nuclear market is expensive for the construction cost (2.0 cent per kWh) but cheap for the fuel cost (1.4 cent per kWh). The introduction of LWR can reduce gross amounts of fuels and then increase the gross domestic products (GDP). (3) During a nuclear cycle the emitted carbons from LWR (22g per kWh) is from one 23rd to one 44th of those from fossil power plants. The gross electricity produced in Japan in 2004 is about 8,651 TWh. Emitted carbons assuming that coal and petroleum are main carbon contributors are 7.43E08 ton. The no fossil fuels can suppress the amounts by 3.79E08 ton, where the contributing ratio of nuclear energy is 57%. An indirect effect of green technology by Japanese LWRs is estimated to be from 3,993 to 5,989 million dollars.