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Shimodaira, Masaki; Tobita, Toru; Takamizawa, Hisashi; Katsuyama, Jinya; Hanawa, Satoshi
Proceedings of ASME 2020 Pressure Vessels and Piping Conference (PVP 2020) (Internet), 7 Pages, 2020/08
In JEAC 4206 which prescribes the methodology for assessing the structural integrity of reactor pressure vessels (RPVs), an under-clad crack (UCC) at the inner surface of RPV is postulated, and it is required that the fracture toughness of RPV steels is higher than stress intensity factor for at the crack tip during the pressurized thermal shock event. In the present study, to investigate the effect of cladding on the fracture toughness, we performed three-point bending fracture toughness tests and finite element analyses (FEAs) for an RPV steel containing an UCC or a surface crack, and the constraint effect for UCC was also discussed. As the result, we found that the fracture toughness for UCC was considerably higher than that for surface crack. On the other hand, the FEAs showed that the cladding decreased the constraint effect for UCC.
Goto, Minoru; Okumura, Keisuke; Nakagawa, Shigeaki; Inaba, Yoshitomo; Matsuura, Hideaki*; Nakaya, Hiroyuki*; Katayama, Kazunari*
Fusion Engineering and Design, 136(Part A), p.357 - 361, 2018/11
Times Cited Count：6 Percentile：52.24(Nuclear Science & Technology)A High Temperature Gas-cooled Reactor (HTGR) is proposed as a tritium production device, which has the potential to produce a large amount of tritium using 
Li(n,
)T reaction. In the HTGR design, generally, boron is loaded into the core as a burnable poison to suppress excess reactivity. In this study, lithium is loaded into the HTGR core instead of boron and is used as a burnable poison aiming to produce thermal energy and tritium simultaneously. The nuclear characteristics and the fuel temperature were calculated to confirm the feasibility of the lithium-loaded HTGR. It was shown that the calculation results satisfied the design requirements and hence the feasibility was confirmed for the lithium-loaded HTGR, which produce thermal energy and tritium.
Takamatsu, Kuniyoshi; Nakagawa, Shigeaki
Nihon Genshiryoku Gakkai Wabun Rombunshi, 5(1), p.45 - 56, 2006/03
The HTTR (High Temperature Engineering Test Reactor), which has thermal output of 30MW, coolant inlet temperature of 395
C and coolant outlet temperature of 850C/950C, is a first high temperature gas-cooled reactor (HTGR) in Japan. The HTGR has a high inherent safety potential to accident condition. Safety demonstration tests using the HTTR are underway in order to demonstrate such excellent inherent safety features of the HTGR. A one-point core dynamics approximation with one fuel channel model had applied to this analysis. It was found that the analytical model for core dynamics couldn't simulate the reactor power behavior accurately. This report proposes an original method using temperature coefficients of some regions in the core. It is crucial to evaluate this method precisely to simulate a performance of HTGR during the test.
Maebara, Sunao; Goniche, M.*; Kazarian, F.*; Seki, Masami; Ikeda, Yoshitaka; Imai, Tsuyoshi*; Beaumont, B.*
Review of Scientific Instruments, 76(5), p.053501_1 - 053501_7, 2005/05
Times Cited Count：1 Percentile：10.09(Instruments & Instrumentation)Development of a plasma facing module using Cold Isostatic Pressing Graphite (CIPG) had been done for a heat-resistant LHCD antenna. A thin stainless film (10
m), molybdenum film (10m) and copper film (50m) are laid to overlap each other on the CIPG materials, the CIPG surfaces were successfully coated with copper layer by diffusion bonding method. This module has four waveguides and a water cooling channel, the length is 206 mm. High power long pulse operation was successfully achieved up to 250 kW (125 MW/m
)/700s. The module has been successfully tested at a RF power density which is equivalent, in terms of RF electric field (5kV/cm), to the one proposed for the LHCD antenna of ITER-FEAT. The outgassing rate of the copper-coated CIPG is estimated to be 3.2-5.1
10
Pa.m
/s.m at 100C, it is assessed that a pumping system is not required to evacuate the pressure in the LHCD antenna.
Nakagawa, Shigeaki; Sakaba, Nariaki; Takamatsu, Kuniyoshi; Takada, Eiji*; Tochio, Daisuke; Owada, Hiroyuki*
JAERI-Tech 2005-015, 26 Pages, 2005/03
JAERI-Tech-2005-015.pdf:1.77MB
Safety demonstration tests using the HTTR are in progress since 2002 to verify the inherent safety features and to improve the safety design and evaluation technologies for HTGRs, as well as to contribute to not only the commercial HTGRs but also the research and development for the VHTR one of the Generation IV reactor candidates. This paper describes the reactivity insertion test (SR-3), the coolant flow reduction test by tripping of gas circulators (S1C-3/S2C-3), and the partial flow loss of coolant test (SF-2) planned in March 2005 with their detailed test method, procedure and results of pre-test analysis. From the analytical results, it was found that the negative reactivity feedback effect of the core brings the reactor power safely to a stable level without a reactor scram.
Nakagawa, Shigeaki; Tachibana, Yukio; Takamatsu, Kuniyoshi; Ueta, Shohei; Hanawa, Satoshi
Nuclear Engineering and Design, 233(1-3), p.291 - 300, 2004/10
Times Cited Count：8 Percentile：48.76(Nuclear Science & Technology)The High Temperature Gas-cooled Reactor (HTGR) is particularly attractive due to its capability of producing high temperature helium gas and due to its inherent safety characteristics. The High Temperature Engineering Test Reactor (HTTR), which is the first HTGR in Japan, was successfully constructed at the Oarai Research Establishment of the Japan Atomic Energy Research Institute. The HTTR achieved full power of 30MW at a reactor outlet coolant temperature of about 850C on December 7, 2001 during the "rise-to-power tests". Two kinds of tests were carried out during the "rise-to-power tests". One is commissioning test to get operation permit by the government and another is test to confirm a performance of the reactor, heat exchanger, control system. From the test results of the "rise-to-power tests" up to 30MW, the functionality of the reactor and the cooling system were confirmed, and it was also confirmed that an operation of reactor facility can be performed safely.
Nakagawa, Shigeaki; Takamatsu, Kuniyoshi; Tachibana, Yukio; Sakaba, Nariaki; Iyoku, Tatsuo
Nuclear Engineering and Design, 233(1-3), p.301 - 308, 2004/10
Times Cited Count：22 Percentile：79.05(Nuclear Science & Technology)Safety demonstration tests using the High Temperature Engineering Test Reactor (HTTR) are conducted for demonstrating inherent safety features of High Temperature Gas-cooled Reactors (HTGRs) as well as for providing core and plant transient data for validation of HTGR safety analysis codes. The safety demonstration tests are divided to the first phase and second phase tests. In the first phase tests, simulation tests of anticipated operational occurrences and anticipated transients without scram (ATWS) are conducted. The second phase tests will simulate accidents such as a depressurization accident (loss of coolant accident). The first phase tests simulating reactivity insertion events and coolant flow reduction events started in FY 2002. The first phase safety demonstration tests will continue until FY 2005, and the second phase tests will be carried out from FY 2006.
Onuki, Akira; Takase, Kazuyuki; Kureta, Masatoshi; Yoshida, Hiroyuki; Tamai, Hidesada; Liu, W.; Akimoto, Hajime
Nihon Kikai Gakkai 2003-Nendo Nenji Taikai Koen Rombunshu, Vol.3, p.247 - 248, 2003/08
We start R&D project to develop the predictable technology for thermal-hydraulic performance of Reduced-Moderation Water Reactor (RMWR) in collaboration with power company/reactor vendor/university since 2002. The RMWR can attain the favorable characteristics such as effective utilization of uranium resources, multiple recycling of plutonium, high burn-up and long operation cycle, based on matured BWR technologies. MOX fuel assemblies with tight lattice arrangement are used to increase the conversion ratio by reducing the moderation of neutron energy. Increasing the in-core void fraction also contributes to the reduction of neutron moderation. The confirmation of thermal-hydraulic feasibility is one of the most important R&D items for the RMWR because of the tight lattice configuration. This series presentation focuses on the feasibility study and shows the R&D plan using large-scale test facility and advanced numerical simulation technology.
Kuroda, Toshimasa*; Hatano, Toshihisa; Miki, Nobuharu*; Hiroki, Seiji; Enoeda, Mikio; Omori, Junji*; Sato, Shinichi*; Akiba, Masato
JAERI-Tech 2002-098, 136 Pages, 2003/02
JAERI-Tech-2002-098.pdf:24.33MB
no abstracts in English
Murakami, Yoshiki*; Amano, Tsuneo*; Shimizu, Katsuhiro; Shimada, Michiya; Ogawa, Yuichi*
JAERI-Research 2001-049, 58 Pages, 2001/11
JAERI-Research-2001-049.pdf:3.11MB
no abstracts in English
Iigaki, Kazuhiko; Sakaba, Nariaki; Kawaji, Satoshi; Iyoku, Tatsuo
Transactions of 16th International Conference on Structural Mechanics in Reactor Technology (SMiRT-16) (CD-ROM), 7 Pages, 2001/08
no abstracts in English
Murakami, Yoshiki*; Senda, Ikuo; Chudnovskiy, A.*; Vayakis, G.*; Polevoi, A. R.*; Shimada, Michiya
Purazuma, Kaku Yugo Gakkai-Shi, 73(7), p.712 - 729, 2001/07
no abstracts in English
Iwai, Yasunori; Yamanishi, Toshihiko; Nishi, Masataka
JAERI-Tech 2001-027, 29 Pages, 2001/03
JAERI-Tech-2001-027.pdf:1.11MB
no abstracts in English
Nakagawa, Shigeaki; Saikusa, Akio; Kunitomi, Kazuhiko
Nuclear Technology, 133(2), p.141 - 152, 2001/02
Times Cited Count：3 Percentile：27.07(Nuclear Science & Technology)no abstracts in English
Naka Fusion Research Establishment
JAERI-Review 2000-030, 113 Pages, 2001/01
JAERI-Review-2000-030.pdf:10.11MB
no abstracts in English
Ebisawa, Katsuyuki*; Costley, A.*; Donn
, A. J. H.*; Janeschitz, G.*; Kasai, Satoshi; Malaquias, A.*; Vayakis, G.*; Walker, C. I.*; Yamamoto, Shin; Zavariaev, V.*
Review of Scientific Instruments, 72(1), p.545 - 550, 2001/01
Times Cited Count：18 Percentile：67.36(Instruments & Instrumentation)no abstracts in English
Araya, Fumimasa; Kureta, Masatoshi; Akimoto, Hajime
Proceedings of 2nd Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-2), p.309 - 314, 2000/00
no abstracts in English
Onuki, Akira; Yoshida, Hiroyuki; Akimoto, Hajime
Proceedings of ANS International Meeting on Best Estimate Methods in Nuclear Installations Safety Analysis (BE-2000) (CD-ROM), 17 Pages, 2000/00
no abstracts in English
Seki, Yasushi
New Energy Systems and Conversions, p.355 - 359, 1999/00
no abstracts in English
Onuki, Akira; Okubo, Tsutomu; Akimoto, Hajime
Proceedings of 7th International Conference on Nuclear Engineering (ICONE-7) (CD-ROM), 10 Pages, 1999/00
no abstracts in English
