Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Akimoto, Hajime; Sugawara, Takanori
JAEA-Data/Code 2016-008, 87 Pages, 2016/09
JAEA-Data-Code-2016-008.pdf:15.62MB
Thermal hydraulic behavior in a lead-bismuth cooled accelerator driven system (ADS) is analyzed under normal operation condition. Input data for the ADS version of J-TRAC code have been constructed to integrate the conceptual design. The core part of the ADS is modeled in detail to evaluate the core radial power profile effect on the core cooling. As the result of the analyses, the followings are found; (1) Both maximum clad temperature and fuel temperature are below the design limits. (2) The radial power profile has little effect on the coolant flow distribution among fuel assemblies. (3) The radial power profile has little effect on the heat transfer coefficients along fuel rods. (4) The thermal hydraulic behaviors along four steam generators are identical. The thermal hydraulic behaviors along two pumps are also identical. A fast running input data is developed by the simplification of the detailed input data based on the findings mentioned above.
Akimoto, Hajime
JAEA-Data/Code 2014-031, 75 Pages, 2015/03
JAEA-Data-Code-2014-031.pdf:37.23MB
A thermal-hydraulic analysis code for transmutation system with lead-bismuth cooled accelerator-driven system (ADS) has been developed using the Japanese-version of Transient Reactor Analysis Code (J-TRAC) as the framework to apply the design studies of ADS. To identify the required capabilities of the thermal-hydraulic analysis code for ADS, previous thermal-hydraulic analyses of light water reactors, sodium-cooled fast reactor and ADS have been surveyed. To make up for insufficient capabilities of the J-TRAC code as a thermal-hydraulic analysis code of ADS, physical properties of lead-bismuth eutectic (LBE), argon gas and nitride nuclear fuel were implemented to the J-TRAC code. It was confirmed that the implemented capabilities worked as expected through verification calculations on (1) single-phase LBE flow, (2) heat transfer in a fuel assembly, and (3) heat transfer in a steam generator.
Nagatake, Taku; Tamai, Hidesada; Akimoto, Hajime; Yoshida, Hiroyuki; Takase, Kazuyuki
Nihon Kikai Gakkai Dai-25-Kai Keisan Rikigaku Koenkai Rombunshu (CD-ROM), p.718 - 719, 2012/10
no abstracts in English
Akimoto, Hajime
Konsoryu, 26(3), p.266 - 272, 2012/09
Research and development activities for two-phase analysis codes for nuclear reactor design and safety analyses have been reviewed focusing in recent twenty five years. For reactor safety evaluation, large-scale tests were performed to confirm effectiveness of ECCS in 1980's and 1990's. These test results were succeeded to so-called best-estimate codes such as RELAP5, TRAC codes. Severe accident researches were performed in 1980's and 1990's and accident management methods were studied. Detailed simulation methods such as subchannel analysis, multi-dimensional analyses have been developed based on test results and computational technology enhancement in 1990's and 2000's. Future scope is summarized briefly.
Tamai, Hidesada; Akimoto, Hajime; Takase, Kazuyuki
Nihon Genshiryoku Gakkai Wabun Rombunshi, 11(1), p.8 - 12, 2012/03
The Fukushima Daiichi Nuclear Plant Unit 1 accident was investigated with TRAC-BF1 code in order to confirm the effect of an isolation condenser (IC) on core cooling analytically. The analysis shows that it is too late to cool fuel rods, once the core is heated up because of the lack of coolant derived from discharge of steam through a safety relief valve. It also shows that early start-up of the IC is essential to avoid the core meltdown under station blackout conditions.
Akimoto, Hajime
Doryoku, Netsu Shisutemu Handobukku, p.240 - 264, 2010/01
As a chapter of "Handbook of power and thermal systems", outline of nuclear power system is summarized, including (1) reactor type, (2) current status of nuclear power generation, and (3) status of R&D of advanced reactors. It is also explained for major components and brief R&D history of light water reactors, heavy water reactors, gas-cooled reactors, fast breeding reactors and fusion reactors.
Akimoto, Hajime; Ohshima, Hiroyuki; Kamide, Hideki; Nakagawa, Shigeaki; Ezato, Koichiro; Takase, Kazuyuki; Nakamura, Hideo
Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) (CD-ROM), 7 Pages, 2009/09
Thermal-hydraulics researches at JAEA are performed in many nuclear R&D areas that investigate fusion reactor, fast reactors (FR), high temperature gas cooled reactors (HTGR), and light water reactors (LWRs). These researches are composed of both experimental and analytical efforts. Experimental efforts cover from small-scale fundamental works to large-scale system-integrated tests. Analytical efforts cover both so-called one-dimensional system analysis codes and detailed three-dimensional CFD codes. The thermal-hydraulic phenomena dealt at JAEA cover both normal operation conditions and accident conditions including severe accidents in LWR and FR. The phenomena include single phase flow of water, supercritical water, helium and sodium, two-phase flow of steam-water or sodium-argon, and multi-phase flows encountered in severe accidents. In this paper, we will summarize current status and future directions of thermal-hydraulic researches at JAEA.
Yamaji, Akifumi; Takizuka, Takakazu; Nabeshima, Kunihiko; Iwamura, Takamichi; Akimoto, Hajime
Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9366_1 - 9366_8, 2009/05
This study has been carried out in series with the other study, "Criticality of Low Enriched Uranium Fueled Core" to explore the possibilities of a solid reactor electricity generation system for supplying propulsion power of a deep space explorer. The design ranges of three different systems are determined with respect to the electric power, the radiator mass, and the operating temperatures of the heat-pipes and thermoelectric converters. The three systems are the solid thermal conduction system (STC), core surface cooling with heat-pipe system (CSHP), and the core direct cooling with heat-pipe system (CDHP). The evaluated electric powers widely cover the 1 to 100 kW range, which had long been claimed to be the range that lacked the power sources in space. Therefore, the concepts shown by this study may lead to a breakthrough of the human activities in space. The working temperature ranges of the main components, namely the heat-pipes and thermoelectric converters, are wide and cover down to relatively low temperatures. This is desirable from the viewpoints of broadening the choices, reducing the development needs, and improving the reliabilities of the devices. Hence, it is advantageous for an early establishment of the concept.
Kugo, Teruhiko; Akie, Hiroshi; Yamaji, Akifumi; Nabeshima, Kunihiko; Iwamura, Takamichi; Akimoto, Hajime
Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9371_1 - 9371_8, 2009/05
Combining a nuclear reactor with thermoelectric converters is expected to be one of promising options to supply a propulsion power for deep space explorers. One of the key features of the concept is to use low enriched uranium fuels from the viewpoint of nuclear non-proliferation. Fuels of uranium oxide, nitride and metal were examined. Zirconium and yttrium hydrides, beryllium, zirconium beryllide and graphite were considered as moderators. Reflectors of beryllium, beryllium oxide, zirconium beryllide and graphite were taken into consideration. A criticality survey of the core was performed by changing the ratio of the fuel, moderator and structure, and the reflector thickness. As a result from the viewpoint of a smaller mass of reactor, it is better to use thermal spectrum cores than fast ones, and the metal hydride moderators than beryllium or graphite. For example, a combination of uranium nitride, yttrium hydride and beryllium reflector achieves a reactor mass of as low as 500kg.
Akimoto, Hajime; Anoda, Yoshinari; Takase, Kazuyuki; Tamai, Hidesada; Yoshida, Hiroyuki
Genshiryoku Kyokasho "Genshiryoku Netsuryudo Kogaku", 336 Pages, 2009/03
no abstracts in English
Yoshida, Hiroyuki; Onuki, Akira; Misawa, Takeharu; Takase, Kazuyuki; Akimoto, Hajime
Nuclear Technology, 164(1), p.45 - 54, 2008/10
Times Cited Count：2 Percentile：16.38(Nuclear Science & Technology)Liu, W.; Onuki, Akira; Yoshida, Hiroyuki; Kureta, Masatoshi; Takase, Kazuyuki; Akimoto, Hajime
Heat Transfer Engineering, 29(8), p.704 - 711, 2008/08
Times Cited Count：2 Percentile：16.93(Thermodynamics)With using the research achievements on thermal-hydraulic characteristics so far derived for the target tight lattice core, this paper studies the thermal feasibility of the designed 1356 MWe FLWR core using a modified transient analysis code TRAC-BF1. The newest critical power correlation developed at JAEA for tight lattice rod bundles is implemented to judge the occurrence of boiling transition from nucleate boiling to film boiling. The pressure drop in two-phase flow region is evaluated by Martinelli-Nelson two-phase multiplier. In the analyses, the 900 fuel assemblies in the designed core are modeled into 12 fuel channels according to the relative mass and power distributions. Analyses to the postulated abnormal transient events that may be possibly met in the operation of the FLWR are performed with 
MCPRs being evaluated. The necessary coolant flow rate then is calculated based on the evaluated MCPRs. As the results, for a natural circulation type FLWR, the operation limited MCPR is 1.19. For a forced circulation type FLWR, the one is 1.32.
Tamai, Hidesada; Kureta, Masatoshi; Liu, W.; Sato, Takashi; Nakatsuka, Toru; Onuki, Akira; Akimoto, Hajime
Journal of Nuclear Science and Technology, 45(6), p.567 - 574, 2008/06
Times Cited Count：1 Percentile：9.73(Nuclear Science & Technology)The confirmation of thermal-hydraulic performance is one of the most important requirements for the design of the FLWR. A large-scale thermal-hydraulic experiment using a tight-lattice 37-rod bundle test section with a bowed rod was carried out with pressure ranging from 2-9 MPa and mass velocity at 200-1000 kg/(m
s). It was confirmed that boiling transition (BT) occurs downstream of the rod contact point, and that the wall temperature trace during the BT follows the typical BT pattern of BWR. Critical power with a bowed rod is about 10 percent lower than that without rod bowing. The critical power increases monotonically with increase in mass velocity, with decrease in inlet water temperature, and with decrease in exit pressure, and these trends are similar to those of the basic bundle without rod bowing. Thus, there is negligible effect of rod bowing on the dependence of critical power on the mass velocity, the inlet temperature and the exit pressure.
Kureta, Masatoshi; Yoshida, Hiroyuki; Tamai, Hidesada; Onuki, Akira; Akimoto, Hajime
Konsoryu Kenkyu No Shinten, 3, p.99 - 109, 2008/06
An estimation of void fraction in tight-lattice rod bundles was carried out. Five types of void fraction experiments with 7-, 14-, 19- and 37-rod and rod-gap of 1.0 - 1.3 mm bundle and spacer effect tests were conducted ranging from 0.1 to 7.2 MPa. Extensibility of a TRAC-BF1 code and one-dimensional drift-flux model to the tight-lattice rod bundle was studied. The TRAC-BF1 and the model calculated the void fraction with good agreement to data in case of relatively high quality and void fraction region. Applicability of a NASCA, ACE-3D, TPFIT codes to the tight-lattice rod bundle was verified by comparing with the three-dimensional void fraction data measured by neutron tomography. Tendency of the calculated void fraction by these codes and measured data was similar. Vapor distribution and velocity profile of water and vapor were discussed based on data.
Izawa, Yasukazu*; Nishihara, Katsunobu*; Tanuma, Hajime*; Sasaki, Akira; Murakami, Masakatsu*; Sunahara, Atsushi*; Nishimura, Hiroaki*; Fujioka, Shinsuke*; Aota, Tatsuya*; Shimada, Yoshinori*; et al.
Journal of Physics; Conference Series, 112, p.042047_1 - 042047_4, 2008/00
Times Cited Count：9 Percentile：93.98(Physics, Fluids & Plasmas)In the development of a high power EUV source used in the EUV lithography system, we have been constructed EUV database of laser-produced tin plasma by the theoretical and experimental studies. On the basis of our understanding, the optimum conditions of lasers and plasmas were clarified, and we proposed the guidelines of laser plasma to obtain clean, efficient and high power EUV source for the practical EUV lithography system. In parallel to such studies, novel targets and high power laser system to generate the optimized EUV source plasma have been developed.
Yoshida, Hiroyuki; Nagayoshi, Takuji*; Takase, Kazuyuki; Akimoto, Hajime
Journal of Power and Energy Systems (Internet), 2(1), p.250 - 258, 2008/00
Misawa, Takeharu; Yoshida, Hiroyuki; Akimoto, Hajime
Journal of Power and Energy Systems (Internet), 2(1), p.262 - 270, 2008/00
Onuki, Akira; Kureta, Masatoshi; Yoshida, Hiroyuki; Tamai, Hidesada; Liu, W.; Misawa, Takeharu; Takase, Kazuyuki; Akimoto, Hajime
Journal of Power and Energy Systems (Internet), 2(1), p.229 - 239, 2008/00
Kureta, Masatoshi; Tamai, Hidesada; Yoshida, Hiroyuki; Onuki, Akira; Akimoto, Hajime
Journal of Power and Energy Systems (Internet), 2(1), p.271 - 282, 2008/00
An estimation of void fraction in a tight-lattice rod bundle was needed of the design of the Innovative Water Reactor for Flexible Fuel Cycle (FLWR). For this purpose, we measured the void fraction and studied the behaviors of boiling flow. The void fraction was measured by a neutron radiography, a quick-shut-valve technique, and an electro void fraction meter. The data were taken using the 7-, 14-, 19- and 37-rod bundle test sections with the rod gap of 1.0 or 1.3 mm under from atmospheric pressure to 7.2 MPa. Followings were made clear: (1) numerical analysis codes calculate the similar distribution to the data, and (2) TRAC-BF1 code and drift-flux model tends to overestimate the void fraction at lower quality region.
Liu, W.; Tamai, Hidesada; Kureta, Masatoshi; Onuki, Akira; Akimoto, Hajime
Journal of Power and Energy Systems (Internet), 2(1), p.240 - 249, 2008/00
This paper describes the critical power characteristics in a 37-rod tight-lattice bundle with rod bowing under transient states. It is observed that transient Boiling Transition (BT) always occurs axially at exit elevation of upper high-heat-flux region and transversely in the central area of the bundle, which is same as that under steady state. For the postulated power increase and flow decrease cases that may be possibly met in a normal operation of the FLWR, it is confirmed that no BT occurs when Initial Critical Power Ratio (ICPR) is 1.3. Moreover, when the transients are run under severer ICPR that causes BT, the transient critical powers are generally same as the steady ones. The experiments are analyzed with TRAC-BF1 code. The code shows good prediction for the occurrence or the non occurrence of the BT and predicts the BT starting time conservatively. Traditional quasi-steady state prediction of the transient BT is confirmed being applicable for the postulated abnormal transient processes in the tight-lattice bundle with rod bowing.
