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Sagayama, Yutaka; Ando, Masato
Nihon Genshiryoku Gakkai-Shi ATOMO
, 60(3), p.162 - 167, 2018/03
The Generation IV international Forum (GIF) has led international collaborative efforts to develop six next generation nuclear energy systems, such as Sodium-cooled Fast Reactor (SFR), Lead-cooled Fast Reactor (LFR), Gas-cooled Fast Reactor (GFR), Molten Salt Reactor (MSR), Supercritical Water-cooled Reactor (SCWR), and Very High Temperature Reactor (VHTR), which have superior characteristics for the Safety and Reliability, Economics, Sustainability, Proliferation Resistance and Physical Protection. Some systems are already in the Demonstration Phase and the commercialization of the system in 2030s, which is the target of GIF, comes into sight.
Tatsumoto, Hideki; Kato, Takashi; Aso, Tomokazu; Hasegawa, Shoichi; Ushijima, Isamu*; Otsu, Kiichi*; Ikeda, Yujiro
LA-UR-06-3904, Vol.2, p.426 - 434, 2006/06
In JSNS, Cadmium has been selected as a poison material in a hydrogen moderator to obtain narrow neutron pulse. The concern to adopt to Cd is how to bond Cd and Al alloy plate. R&Ds for bonding have been performed. But good bonding has not been obtained. Consequently, heat transfer between Cd poison and cryogenic hydrogen was studied for the case of insufficient bonding. The heat transfers for various bonding ratios were analyzed by CFD code (STAR-CD) without any turbulence model. The temperature rise in Cd poison for insufficient bonding was estimated. As a result, even the case of the bonding ratio of only 5 %, the maximum temperature of Cd is around 75K. Therefore, the expected heat transfer between the Cd poison and the hydrogen should be sufficient for insufficient bonding. Then, it is found that the any bonding method should be available for manufacturing method of Cd poison.
Tatsumoto, Hideki; Kato, Takashi; Aso, Tomokazu; Ushijima, Isamu*; Hasegawa, Shoichi; Otsu, Kiichi*
JAERI-Tech 2005-019, 16 Pages, 2005/03
JAERI-Tech-2005-019.pdf:1.29MB
As one of the main experimental facilities in J-PARC, an intense spallation neutron source (JSNS) is constructed. In JSNS, cryogenic hydrogen with temperature of 20 K and pressure of 0.5 to 1.5 MPa was selected as the moderator. The total nuclear heating at the moderators is estimated to be 3.7 kW for proton beam power of 1 MW. A cryogenic hydrogen circulation system, which plays a role in cooling spallation neutron and moderators, has been designed. For a certain operation condition, it is possible to occur boiling in the moderators. The boiling phenomenon would have an influence on the neutronic performance and the safety of the moderators. The heat transfer mechanism of cryogenic hydrogen in the moderators needs to be estimated. However, the mechanism has not been clarified until now. In this paper, the heat transfer of cryogenic hydrogen was estimated by using properties of cryogenic hydrogen and the heat transfer correlations used in other fluids, and then the operation condition of the cryogenic hydrogen system has been considered.
Enoeda, Mikio; Kosaku, Yasuo; Hatano, Toshihisa; Kuroda, Toshimasa*; Miki, Nobuharu*; Homma, Takashi; Akiba, Masato; Konishi, Satoshi; Nakamura, Hirofumi; Kawamura, Yoshinori; et al.
Nuclear Fusion, 43(12), p.1837 - 1844, 2003/12
Times Cited Count：103 Percentile：93.16(Physics, Fluids & Plasmas)no abstracts in English
Kurihara, Ryoichi; Watanabe, Kenichi*; Konishi, Satoshi
JAERI-Review 2003-020, 37 Pages, 2003/07
JAERI-Review-2003-020.pdf:2.08MB
no abstracts in English
Kosaku, Yasuo; Yanagi, Yoshihiko*; Enoeda, Mikio; Akiba, Masato
Fusion Science and Technology, 41(3), p.958 - 961, 2002/05
As a candidate DEMO blanket, the design of solid breeder blanket cooled by supercritical water has been performed. The candidate structural material is F82H. The coolant is supercritical water (pressure; 25 MPa, temperature; 550-780K) to achieve high generation efficiency. The temperature of cooling tubes in tritium breeder zone has been evaluated at 650-800K. In this temperature range, tritium permeation must be investigated from the view point of safety management, because high temperature coolant is directly supplied to the power generation system. In the present work, the tritium permeation into the first wall cooling water by the implantation and that through cooling tubes in tritium breeder zone have been evaluated. Assuming tritium injection energy and flux are same as SSTR, the calculated value of the tritium permeation rate into the first wall cooling water is 68.3 g/day. On the other hand, that of the permeation rate through cooling tubes is 75.3 g/day (20% of generated tritium) when helium gas flows so that tritium partial pressure becomes 1 Pa at the outlet.
Yanagi, Yoshihiko*; Sato, Satoshi; Enoeda, Mikio; Hatano, Toshihisa; Kikuchi, Shigeto*; Kuroda, Toshimasa*; Kosaku, Yasuo; Ohara, Yoshihiro
Journal of Nuclear Science and Technology, 38(11), p.1014 - 1018, 2001/11
Times Cited Count：25 Percentile：83.65(Nuclear Science & Technology)no abstracts in English
Nakatsuka, Toru; Oka, Yoshiaki*; Koshizuka, Seiichi*
Nuclear Technology, 134(3), p.221 - 230, 2001/06
Times Cited Count：17 Percentile：74.11(Nuclear Science & Technology)no abstracts in English
Koshizuka, Seiichi*
JNC TJ9400 2000-011, 102 Pages, 2000/03
JNC-TJ9400-2000-011.pdf:2.71MB
In order to evaluate the possibility to achieve high electric power by a fast reactor with supercritical light water, the design study was carried out on a large fast reactor core with high coolant outlet temperature (SCFR-H). Since the reactor coolant circuit uses once-through direct cycle where all feedwater flows through the core to the turbine at supercritical pressure, it is possible to design much simpler and more compact reactor systems and to achieve higher thermal efficiency than those of current light water reactors. The once-through direct cycle system is employed in current fossil-fired power plants. In the present study, three types of core were designed. The first is SCFR-H with blankets cooled by ascending flow, the second is SCFR-H with blankets cooled by descending flow and the third is SCFR-H with high thermal power. Every core was designed to achieve the thermal efficiency over 43%, positive coolant density reactivity coefficient and electric power over 1600MW. Core characteristics of SCFR-Hs were compared with those of SCLWR-H (electric power: 1212MW), which is a thermal neutron spectrum reactor cooled and moderated by supercritical light water, with the same diameter of the reactor pressure vessel. It was shown that SCFR-H could increase the electric power about l.7 times maximally. From the standpoint of the increase of a reactor thermal power, a fast reactor has advantages as compared with a thermal neutron reactor, because it can increase the power density by adopting tight fuel lattices and eliminating the moderator region. Thus, it was concluded that a reactor cooled by supercritical light water could further improve the cost competitiveness by using a fast neutron spectrum and achieving a higher thermal power.
Nakatsuka, Toru; Oka, Yoshiaki*; Koshizuka, Seiichi*
Proceedings of 8th International Conference on Nuclear Engineering (ICONE-8) (CD-ROM), p.9 - 0, 2000/00
no abstracts in English
; Ochiai, Masaaki
JAERI-Conf 98-013, 279 Pages, 1998/09
no abstracts in English
