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Takeda, Tetsuaki*; Inagaki, Yoshiyuki; Aihara, Jun; Aoki, Takeshi; Fujiwara, Yusuke; Fukaya, Yuji; Goto, Minoru; Ho, H. Q.; Iigaki, Kazuhiko; Imai, Yoshiyuki; et al.
High Temperature Gas-Cooled Reactors; JSME Series in Thermal and Nuclear Power Generation, Vol.5, 464 Pages, 2021/02
As a general overview of the research and development of a High Temperature Gas-cooled Reactor (HTGR) in JAEA, this book describes the achievements by the High Temperature Engineering Test Reactor (HTTR) on the designs, key component technologies such as fuel, reactor internals, high temperature components, etc., and operational experience such as rise-to-power tests, high temperature operation at 950
C, safety demonstration tests, etc. In addition, based on the knowledge of the HTTR, the development of designs and component technologies such as high performance fuel, helium gas turbine and hydrogen production by IS process for commercial HTGRs are described. These results are very useful for the future development of HTGRs. This book is published as one of a series of technical books on fossil fuel and nuclear energy systems by the Power Energy Systems Division of the Japan Society of Mechanical Engineers.
Osaki, Hirotaka; Shimazaki, Yosuke; Sumita, Junya; Shibata, Taiju; Konishi, Takashi; Ishihara, Masahiro
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 8 Pages, 2015/05
For the design on the VHTR graphite components, it is desirable to employ graphite material with higher strength. IG-430 graphite has been developed as an advanced candidate for VHTR. However, the new developed IG-430 does not have enough databases for the design of HTGR. In this paper, the compressive strength (Cs) of IG-430, one of important strengths for design data, is statistically evaluated. The component reliability is evaluated based on the safety factors defined by the graphite design code, and the applicability as the VHTR graphite material is discussed. It was found that IG-430 has higher strength (about 11%) and lower standard deviation (about 27%) than IG-110 which is one of traditional graphites used for HTGR, because the crack in IG-430 would not easy to propagate rather than IG-110. Since fracture probability for IG-430 is low, the higher reliability of core-component will be achieved using IG-430. It is expected that IG-430 is applicable for VHTR graphite material.
C continuous operation of HTTRSumita, Junya; Shimazaki, Yosuke; Shibata, Taiju
Journal of Nuclear Science and Technology, 51(11-12), p.1364 - 1372, 2014/11
Times Cited Count：7 Percentile：46.64(Nuclear Science & Technology)Graphite material is used for internal structures in High Temperature Gas-cooled Reactor (HTGR). The core components and graphite core support structures are so designed as to maintain the structural integrity to keep core cooling capability. In order to confirm that the core components and graphite core support structures satisfy the design requirement, the temperatures of the reactor internals are measured during the reactor operation. Surveillance test of graphite specimens and in-service inspection (ISI) using TV camera are planned in conjunction with the refueling. This paper describes the evaluation results of the integrity of the core components and graphite core support structures during the high temperature 950C continuous operation, reactor outlet temperature of 950C for 50 days, in HTTR (High Temperature Engineering Test Reactor). The design requirements of the core components and graphite core support structure were satisfied during the high temperature 950C continuous operation. The dimensional change of graphite which directly influences the temperature of coolant was evaluated with considering the temperature profiles of fuel block. The magnitude of irradiation-induced dimensional change was about 1.2 times larger than that of isotherm for 1000C. In addition, the programs of surveillance test and ISI using TV camera were introduced.
Takahashi, Hiroki; Kano, Tadashi*; Takahashi, Teruo*; Kyoya, Masahiko; Shimazaki, Junya*
JAERI-Tech 2001-068, 78 Pages, 2001/10
JAERI-Tech-2001-068.pdf:82.36MB
JAERI had carried out design study about a lightweight and compact integral type reactor (an advanced marine reactor) with passive safety equipment, and completed an engineering design. To confirm the design performance and operation performance and to utilize the study of automation of the reactor operation, we have developed a simulator for the integral type reactor. This can be used also for research and development of a small reactor.However, in using for research and development, the improvement in a performance of hardware and a human machine interface of software of the simulator were needed. Renewal of hardware and improvement of software have been conducted. Thereby, the operability of an integral-reactor simulator has been improved. Moreover, versatility, maintainability, extendibility, and the improvement of system performed by using the conventional hardware and software.This is a report on hardware renewal and the interface improvement of the integral type reactor simulator mainly focused on contents of the enhancement in a human machine interface.
Ura, Tamaki*; Takamasa, Tomoji*; Nishimura, Hajime*; Aoki, Taro*; Ueno, Michio*; Maeda, Toshio*; Nakamura, Masato*; Shimazu, Shunsuke*; Tokunaga, Sango*; Shibata, Yozo*; et al.
JAERI-Tech 2001-049, 154 Pages, 2001/07
JAERI-Tech-2001-049.pdf:11.24MB
JAERI has studied on design and operation of a nuclear powered submersible research vessel, which will navigate under sea in the Arctic Ocean, as a part of the design study of advanced marine reactors. This report describes operation conditions and an operating system of the vessel those were discussed by the specialists of hull design, sound positioning, ship motions and oceanography, etc. The design conditions on ship motions for submersible vessels were surveyed considering regulations in our country, and ship motions were evaluated assuming the observation activities in the Arctic Ocean. A submarine transponder system and an on ice communication buoy system were examined as a positioning and communication system supposing the activity under ice. Procedures to secure safety of nuclear powered submersible research vessel were discussed based on the investigation of accidents. These results were reflected to the concept of the nuclear powered submersible research vessel, and subjects fto be settled in the next step were clarified.
Yabuuchi, Noriaki; Takahashi, Masao*; Nakazawa, Toshio; Sato, Kazuo*; Shimazaki, Junya; Ochiai, Masaaki
JAERI-Research 2000-064, 76 Pages, 2001/02
JAERI-Research-2000-064.pdf:5.88MB
no abstracts in English
Yabuuchi, Noriaki; Takahashi, Masao*; Nakazawa, Toshio; Sato, Kazuo*; Shimazaki, Junya; Ochiai, Masaaki
JAERI-Research 2000-063, 69 Pages, 2001/02
JAERI-Research-2000-063.pdf:4.41MB
no abstracts in English
Takahashi, Teruo; Shimazaki, Junya; Nakazawa, Toshio; Yabuuchi, Noriaki; Fukuhara, Yoshifumi*; Kusunoki, Tsuyoshi; Ochiai, Masaaki
JAERI-Tech 2000-039, p.94 - 0, 2000/03
no abstracts in English
Nakazawa, Toshio; Yabuuchi, Noriaki; Takahashi, Hiroki; Shimazaki, Junya
JAERI-Tech 99-008, 45 Pages, 1999/02
no abstracts in English
Yabuuchi, Noriaki; Nakazawa, Toshio; Takahashi, Hiroki; Shimazaki, Junya; Hoshi, Tsutao
JAERI-Tech 97-057, 54 Pages, 1997/11
no abstracts in English
Shimazaki, Junya; Ochiai, Masaaki; Ishida, Toshihisa; Hoshi, Tsutao
10th Pacific Basin Nuclear Conf. (10-PBNC), p.828 - 833, 1996/00
no abstracts in English
Hayashi, Koji; Shimazaki, Junya; Nabeshima, Kunihiko; Shinohara, Yoshikuni; Inoue, Kimio*;
JAERI-Research 95-015, 172 Pages, 1995/03
JAERI-Research-95-015.pdf:4.53MB
no abstracts in English
optimal linear filterSuzuki, Katsuo; Shimazaki, Junya;
Nuclear Science and Engineering, 119, p.128 - 138, 1995/02
Times Cited Count：8 Percentile：62.85(Nuclear Science & Technology)no abstracts in English
Hayashi, Koji; Shimazaki, Junya; Nabeshima, Kunihiko; Shinohara, Yoshikuni; Inoue, Kimio*;
JAERI-Research 95-004, 178 Pages, 1995/01
JAERI-Research-95-004.pdf:4.66MB
no abstracts in English
Hayashi, Koji; Shimazaki, Junya; Shinohara, Yoshikuni*
SMORN-VII,Symp. on Nuclear Reactor Surveillance and Diagnostics,Vol. 1, 0, P. 3_5, 1995/00
no abstracts in English
estimation theorySuzuki, Katsuo; Shimazaki, Junya; Shinohara, Yoshikuni
Nihon Genshiryoku Gakkai-Shi, 36(1), p.79 - 88, 1994/01
Times Cited Count：2 Percentile：27.89(Nuclear Science & Technology)no abstracts in English
Hayashi, Koji; Shimazaki, Junya; Nabeshima, Kunihiko; Shinohara, Yoshikuni; Inoue, Kimio*;
JAERI-M 93-194, 163 Pages, 1993/10
no abstracts in English
Shimazaki, Junya; Ishikawa, Nobuyuki
JAERI-M 93-063, 30 Pages, 1993/05
no abstracts in English
optimal estimator to reactivity estimation of fast breeder reactorSuzuki, Katsuo; Shimazaki, Junya
JAERI-M 93-062, 28 Pages, 1993/03
no abstracts in English
optimal estimator for Linear time-invariant system; State space design approach and its application to reactivity estimationSuzuki, Katsuo; Shimazaki, Junya; Shinohara, Yoshikuni
JAERI-M 93-040, 25 Pages, 1993/03
no abstracts in English
