The Development status of Generation IV reactor systems, 1; Overview

Sagayama, Yutaka; Ando, Masato

Nihon Genshiryoku Gakkai-Shi ATOMO, 60(3), p.162 - 167, 2018/03

https://doi.org/10.3327/jaesjb.60.3_162

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.

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.

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.

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

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