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Journal Articles

Numerical analysis for heat transfer from a Cd poison in cryogenic hydrogen

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.

JAEA Reports

Estimation of heat transfer in moderators filled with cryogenic hydrogen for proton beam power of 1 MW

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.

Journal Articles

Design and technology development of solid breeder blanket cooled by supercritical water in Japan

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:91 Percentile:5.41(Physics, Fluids & Plasmas)

no abstracts in English

JAEA Reports

Literature survey of thermal-hydraulic studies on super-critical pressurized water

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

Journal Articles

Evaluation of tritium permeation in solid breeder blanket cooled by supercritical water

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.

Journal Articles

Nuclear and thermal analyses of supercritical-water-cooled solid breeder blanket for fusion DEMO reactor

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:21 Percentile:16.6(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Startup thermal considerations for supercritical-pressure light water-cooled reactors

Nakatsuka, Toru; Oka, Yoshiaki*; Koshizuka, Seiichi*

Nuclear Technology, 134(3), p.221 - 230, 2001/06

 Times Cited Count:12 Percentile:30.99(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

A Design study of high electric power for fast reactor cooled by super critical light water

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.

Journal Articles

Start-up of superciritical-pressure light water cooled reactors

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

JAEA Reports

The Study Meeting Report on the Undermoderated Spectrum Reactor; March4-5, 1998, JAERI, Tokai Japan

; Ochiai, Masaaki

JAERI-Conf 98-013, 279 Pages, 1998/09

JAERI-Conf-98-013.pdf:11.39MB

no abstracts in English

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