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Journal Articles

Thermochemical decomposition of water

Onuki, Kaoru; Shiozawa, Shusaku

Suiso Enerugi Gijutsu No Tenkai, p.40 - 50, 2010/12

Research and Development on thermochemical water-splitting IS process is reviewed including the principle of thermochemical cycle for hydrogen production from water, the reaction scheme of IS process, and the state-of-the-art in the demonstration of continuous hydrogen production with production rate of 30 litter per an hour, the study on hydriodic acid processing scheme for improving the process heat/mass balance, and the development of corrosion resistant materials and equipments.

Journal Articles

Experimental validation of effectiveness of rod-type burnable poisons on reactivity control in HTTR

Goto, Minoru; Shiozawa, Shusaku; Fujimoto, Nozomu; Nakagawa, Shigeaki; Nakao, Yasuyuki*

Nuclear Engineering and Design, 240(10), p.2994 - 2998, 2010/10

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

In block type high temperature gas-cooled reactors (HTGRs), insertion depth of control rods (CRs) into a core should be retained as shallow as possible to keep fuel temperature below limit through a burnup period. Using burnable poisons (BPs) to control reactivity is considered as a method to resolve this problem as in case of light water reactors (LWRs). BPs design method for LWRs has been validated by experimental data, however, that for HTGRs have not been yet, because there was not burnup characteristics data of HTGRs required for the validation. The High Temperature engineering Test Reactor (HTTR) is a block type HTGRs and it uses BPs to control reactivity. The HTTR has been operated up to middle burnup, and thereby the experimental data was expected to show effect of the BPs on the reactivity control. Hence, in order to validate the BPs design method, we investigated whether the BPs have functioned as designed. As a result, the CRs insertion depth has been retained shallow within allowable range, and then the BPs design method was validated.

Journal Articles

Interpolation and Extrapolation method to analyze irradiation-induced dimensional change data of graphite for design of core components in Very High Temperature Reactor (VHTR)

Shibata, Taiju; Kunimoto, Eiji*; Eto, Motokuni*; Shiozawa, Shusaku; Sawa, Kazuhiro; Oku, Tatsuo*; Maruyama, Tadashi*

Journal of Nuclear Science and Technology, 47(7), p.591 - 598, 2010/07

 Times Cited Count:9 Percentile:56.9(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Fracture stress estimation method of SiC components in IS process

Takegami, Hiroaki; Terada, Atsuhiko; Hino, Ryutaro; Shiozawa, Shusaku

Proceedings of 2nd International Topical Meeting on the Safety and Technology of Nuclear Hydrogen Production, Control and Management (2IST-NH$$_{2}$$) (CD-ROM), p.95 - 100, 2010/06

JAEA has been conducting R&D on thermochemical water-splitting IS process for hydrogen production to meet massive demand in the future hydrogen economy. In the sulfuric acid decomposition process, which is the severest corrosive environment in IS process, silicon carbide (SiC) showed excellent corrosion resistance. Although knowing the strength of the SiC block is important for the reliability assessment and design work, it is difficult to evaluate a large-scale ceramics structure without destructive test. In this study, a novel approach for strength estimation of SiC structure was proposed. A prediction method of minimum strength of the structure was proposed based on effective volume theory and optimized Weibull modulus. The validity of the strength estimation method was verified by destructive test of SiC pipe models, which is simple model of the SiC component. The fracture strength of pipe models satisfied the predicted strength of the model.

JAEA Reports

Draft of standard for graphite core components in High Temperature Gas-cooled Reactor

Shibata, Taiju; Eto, Motokuni*; Kunimoto, Eiji*; Shiozawa, Shusaku; Sawa, Kazuhiro; Oku, Tatsuo*; Maruyama, Tadashi*

JAEA-Research 2009-042, 119 Pages, 2010/01

JAEA-Research-2009-042.pdf:26.28MB

For the design of the graphite components in the High Temperature Engineering Test Reactor (HTTR), the graphite structural design code for the HTTR etc. were applied. However, general standard systems for the High Temperature Gas-cooled Reactor (HTGR) have not been established yet. The authors had studied on the technical issues which is necessary for the establishment of a general standard system for the graphite components in the HTGR. The results of the study were documented and discussed at a "Special committee on research on preparation for codes for graphite components in HTGR" at Atomic Energy Society of Japan (AESJ). As a result, "Draft of Standard for Graphite Core Components in High Temperature Gas-cooled Reactor" was established. In this draft standard, the graphite components are classified three categories (A, B and C) in the standpoints of safety functions and possibility of replacement. For the components in the each class, design standard, material and product standards, and in-service inspection and maintenance standard are determined. This draft standard is the first standard in the world which shows the concept of standard for the graphite core components in HTGR.

JAEA Reports

Expansion of irradiation data by interpolation and extrapolation for design of graphite components in high temperature gas-cooled reactor; Evaluation on IG-110 graphite irradiation data for component design

Kunimoto, Eiji; Shibata, Taiju; Shimazaki, Yosuke; Eto, Motokuni*; Shiozawa, Shusaku; Sawa, Kazuhiro; Maruyama, Tadashi*; Oku, Tatsuo*

JAEA-Research 2009-008, 28 Pages, 2009/06

JAEA-Research-2009-008.pdf:4.6MB

The VHTR is being focused and developed internationally. In Japan, the HTTR of the JAEA is in operation, and research and development for the development of commercial HTGRs are carried out. Nuclear graphites are used for core components of the HTGRs and expansion of irradiation data is necessary when enough irradiation data are not established, because the graphite components in the HTGRs are used at severer condition than that in the HTTR. The necessary database can be established by expansion of existing irradiation data with appropriate interpolation and extrapolation methods. This paper shows the reasonable interpolation and extrapolation method for IG-110 graphite which is used for the HTTR and a major candidate for the VHTR. The interpolation and extrapolation method was developed so as to be general by using the irradiation data of the other graphites. As a result, irradiation properties of the IG-110 graphite were successfully expanded to the VHTR condition for the first time and the irradiation properties being necessary for the design could be developed.

Journal Articles

Present status of HTGR development in Japan

Matsui, Kazuaki*; Shiozawa, Shusaku; Ogawa, Masuro; Yan, X.

Proceedings of 16th Pacific Basin Nuclear Conference (PBNC-16) (CD-ROM), 6 Pages, 2008/10

no abstracts in English

Journal Articles

Application of nuclear energy for activities towards hydrogen society

Shiozawa, Shusaku; Onuki, Kaoru; Inagaki, Yoshiyuki

Denki Kyokai-Ho, (1005), p.18 - 21, 2008/08

no abstracts in English

Journal Articles

Development strategy for non-nuclear grade hydrogen production system coupled with the Japan's HTTR

Sakaba, Nariaki; Sato, Hiroyuki; Ohashi, Hirofumi; Nishihara, Tetsuo; Kunitomi, Kazuhiko; Shiozawa, Shusaku

Proceedings of International Topical Meeting on Safety and Technology of Nuclear Hydrogen Production, Control and Management (ST-NH$$_{2}$$) (CD-ROM), p.355 - 362, 2007/06

Japan Atomic Energy Agency (JAEA) started design studies of the thermochemical water-splitting Iodine Sulphur (IS) process to be coupled with the HTTR to demonstrate hydrogen production from nuclear. It is important from an economic point of view that a non-nuclear grade, rather than nuclear grade, IS process plant be built based on conventional chemical plant construction standards. In order to construct the IS process as a conventional chemical plant, some critical safety issues must been studied and clarified prior to the application for safety case review from the government. JAEA has launched R&D for non-nuclear grade IS process. This paper describes the development strategy for non-nuclear grade hydrogen production system coupled with the Japan's HTTR.

Journal Articles

Basic design and economical evaluation of Gas Turbine High Temperature Reactor 300 (GTHTR300)

Kunitomi, Kazuhiko; Shiozawa, Shusaku; Yan, X.

Proceedings of 2007 International Congress on Advances in Nuclear Power Plants (ICAPP 2007) (CD-ROM), 9 Pages, 2007/05

Japan Atomic Energy Agency had been engaging in the basic design and development of Gas Turbine High Temperature Reactor 300 (GTHTR300). Costs of capital, fuel, and operation and maintenance have been estimated to confirm economical feasibility of the GTHTR300. The cost of electricity for the GTHTR300 is estimated to be below US 3.3 cents/kWh (4 yen/kWh), which is about two-third of that of the current LWRs in Japan. The results confirm that the net power generation cost of the GTHTR300 is much lower than that of the LWR, indicating that the GTHTR300 plant consisting of small-scale reactor units can be economically competitive to the latest large-scale LWR. The contents and results of the economical evaluation will be presented together with the outline of the GTHTR300 design.

Journal Articles

Hydrogen production by using the heat from high-temperature gas-cooled reactors

Shiozawa, Shusaku; Ogawa, Masuro; Hino, Ryutaro; Onuki, Kaoru; Sakaba, Nariaki

Karyoku Genshiryoku Hatsuden, 57(1), p.7 - 12, 2006/01

no abstracts in English

Journal Articles

The HTTR project as the world leader of HTGR research and development

Shiozawa, Shusaku; Komori, Yoshihiro; Ogawa, Masuro

Nihon Genshiryoku Gakkai-Shi, 47(5), p.342 - 349, 2005/05

For the purpose to extend high temperature nuclear heat application, JAERI constructed the HTTR, High Temperature Engineering Test Reactor, and has carried out research and development of high temperature gas cooled reactor system aiming at high efficiency power generation and hydrogen production. This paper explains the history, main results, present status of research and development of HTTR project, international cooperation of research and development of HTGR and future plan aiming at development of Japanese original future HTGR-Hydrogen production system. This paper includes results from the study, which is entrusted from Ministry of Education, Culture, Sports, Science and Technology of Japan.

Journal Articles

Performance test results of mock-up test facility of HTTR hydrogen production system

Ohashi, Hirofumi; Inaba, Yoshitomo; Nishihara, Tetsuo; Inagaki, Yoshiyuki; Takeda, Tetsuaki; Hayashi, Koji; Katanishi, Shoji; Takada, Shoji; Ogawa, Masuro; Shiozawa, Shusaku

Journal of Nuclear Science and Technology, 41(3), p.385 - 392, 2004/03

 Times Cited Count:16 Percentile:70.63(Nuclear Science & Technology)

Prior to construction of a HTTR hydrogen production system, a mock-up test facility was constructed to investigate transient behavior of the hydrogen production system and to establish system controllability. The Mock-up test facility with a full-scale reaction tube is an approximately 1/30 scale model of the HTTR hydrogen production system and an electric heater is used as a heat source instead of a reactor. Before the mock-up test, a performance test of the test facility was carried out in the same pressure and temperature conditions as those of the HTTR hydrogen production system to investigate its performance such as hydrogen production ability, controllability and so on. It was confirmed that hydrogen was stably produced with a hot helium gas about 120Nm$$^{3}$$/h which satisfy the design value and thermal disturbance of helium gas during the start-up could be mitigated within the design value by using a steam generator.

Journal Articles

Status of the Japanese development study of hydrogen production system using HTGR

Shiozawa, Shusaku; Ogawa, Masuro; Inagaki, Yoshiyuki; Onuki, Kaoru; Takeda, Tetsuaki

Proceedings of 18th KAIF-KNS Annual Conference, p.209 - 218, 2003/04

no abstracts in English

Journal Articles

Design study on Gas Turbine High Temperature Reactor (GTHTR300)

Kunitomi, Kazuhiko; Katanishi, Shoji; Takada, Shoji; Takizuka, Takakazu; Nakata, Tetsuo; Yan, X.; Takei, Masanobu; Kosugiyama, Shinichi; Shiozawa, Shusaku

Nihon Genshiryoku Gakkai Wabun Rombunshi, 1(4), p.352 - 360, 2002/12

no abstracts in English

Journal Articles

High temperature gas-cooled reactor

Shiozawa, Shusaku

Nihon Kikai Gakkai Doryoku Enerugi Shisutemu Bumon Nyusu Reta, (24), p.2 - 3, 2002/05

no abstracts in English

Journal Articles

Research and development of HTTR hydrogen production systems

Shiozawa, Shusaku; Ogawa, Masuro; Inagaki, Yoshiyuki; Onuki, Kaoru; Takeda, Tetsuaki; Nishihara, Tetsuo; Hayashi, Koji; Kubo, Shinji; Inaba, Yoshitomo; Ohashi, Hirofumi

Proceedings of 17th KAIF/KNS Annual Conference, p.557 - 567, 2002/04

The research and development program on nuclear production of hydrogen was started on January in 1997 as a study consigned by Ministry of Education, Culture, Sports, Science and Technology. A hydrogen production system connected to the HTTR is being designed to be able to produce hydrogen of about 4000 m3/h by steam reforming of natural gas, using a nuclear heat of 10 MW supplied by the HTTR. In order to confirm controllability, safety and performance of key components in the HTTR hydrogen production system, the facility for an out-of-pile test was constructed on the scale of approximately 1/30 of the HTTR hydrogen production system. Essential tests are also carried out to obtain detailed data for safety review and development of analytical codes. Other basic studies on the hydrogen production technology of thermochemical water splitting called an iodine sulfur (IS) process, has been carried out for more effective and various uses of nuclear heat. This paper describes the present status and a future plan on the R&D of the HTTR hydrogen production systems in JAERI.

Journal Articles

Design of power conversion system of Gas Turbine High Temperature Reactor (GTGTR300)

Takada, Shoji; Takizuka, Takakazu; Kunitomi, Kazuhiko; Yan, X.; Katanishi, Shoji; Kosugiyama, Shinichi; Shiozawa, Shusaku

Nihon Kikai Gakkai Dai-8-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.189 - 192, 2002/00

no abstracts in English

Journal Articles

Safety design philosophy on the Gas Turbine High Temperature Reactor 300 (GTHTR300)

Katanishi, Shoji; Kunitomi, Kazuhiko; Takada, Shoji; Nakata, Tetsuo; Takizuka, Takakazu; Shiozawa, Shusaku

Nihon Kikai Gakkai Dai-8-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.185 - 188, 2002/00

no abstracts in English

Journal Articles

Design study on Gas Turbine High Temperature Reactor (GTHTR300) in FY2001

Kunitomi, Kazuhiko; Katanishi, Shoji; Takada, Shoji; Takizuka, Takakazu; Yan, X.; Nakata, Tetsuo; Takei, Masanobu; Kosugiyama, Shinichi; Shiozawa, Shusaku

Nihon Kikai Gakkai Dai-8-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.181 - 184, 2002/00

no abstracts in English

133 (Records 1-20 displayed on this page)