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Ishii, Katsunori; Aoki, Takeshi; Isaka, Kazuyoshi; Noguchi, Hiroki; Shimizu, Atsushi; Sato, Hiroyuki
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 9 Pages, 2023/05
Nomoto, Yasunobu; Mizuta, Naoki; Morita, Keisuke; Aoki, Takeshi; Okita, Shoichiro; Ishii, Katsunori; Kurahayashi, Kaoru; Yasuda, Takanori; Tanaka, Masato; Isaka, Kazuyoshi; et al.
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 7 Pages, 2023/05
Mizuta, Naoki; Morita, Keisuke; Aoki, Takeshi; Okita, Shoichiro; Ishii, Katsunori; Kurahayashi, Kaoru; Yasuda, Takanori; Tanaka, Masato; Isaka, Kazuyoshi; Noguchi, Hiroki; et al.
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 6 Pages, 2023/05
Aoki, Takeshi; Isaka, Kazuyoshi; Sato, Hiroyuki; Ohashi, Hirofumi
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 7 Pages, 2020/08
The flow distribution analysis performed in the HTGR design has to take into account the interaction thermal and radiation deformations of the graphite structure, and the gaps between the graphite structures forming unintended flow. In the present study, a user-friendly flow network calculation code (FNCC) has been developed on the basis of experiences of High Temperature engineering Test Reactor (HTTR) design for HTGR design with enhanced compatibility with other HTGR design codes and with considering graphite block deformation in iteration process without manual control. The validation of FNCC was performed for the one-column flow distribution test. The analytical results using FNCC showed good agreement with the experimental results. It is concluded that FNCC was validate for the analysis of distributions of flowrate and pressure for the flow network model including the unintended flow paths in prismatic-type HTGRs.
Yamamoto, Tsuyoshi; Hashimoto, Yasunori*; Serizawa, Yasunori*; Inamoto, Shuji*; Sato, Kazuyoshi; Sugie, Tatsuo; Takeuchi, Masaki; Kawano, Yasunori
Fusion Engineering and Design, 89(5), p.532 - 535, 2014/05
Times Cited Count：1 Percentile：8.88(Nuclear Science & Technology)The diagnostic systems are essential for the plasma control and physics understandings. JAEA has proposed the new concept of supervisory system which manages operation sequences, current state and configuration parameters for the measurement based on our experiences in operating plasma diagnostic systems. We designed the supervisory system satisfying the requirements from both CODAC system and diagnostic systems. In our design, the tool which converts operational steps described as flowcharts into EPICS Records templates is introduced. This tool will ensure reduction of the system designers' efforts. We also designed the sequencing simulator that can submit transition commands internally instead of CODAC system for the calibration and commissioning. The mechanism which changes the limit values and consistency check algorithms in accordance with the conditions of the diagnostics system is also proposed.
Ohashi, Hirofumi; Sato, Hiroyuki; Tazawa, Yujiro; Aihara, Jun; Nomoto, Yasunobu; Imai, Yoshiyuki; Goto, Minoru; Isaka, Kazuyoshi; Tachibana, Yukio; Kunitomi, Kazuhiko
JAEA-Technology 2013-017, 71 Pages, 2014/02
JAEA-Technology-2013-017.pdf:3.64MB
Japan Atomic Energy Agency (JAEA) has started a conceptual design of a 50 MWt small-sized high temperature gas cooled reactor (HTGR) for steam supply and electricity generation (HTR50S). Though the safety design of HTR50S was determined based on that of the High Temperature Engineering Test Reactor (HTTR) for the early deployment of HTR50S, the shutdown cooling system, which is the forced cooling heat removal system, was categorized as non-safety class to optimize the protection to provide the highest level of safety that can reasonably be achieved, and the vessel cooling system, which is categorized as the safety class system, was designed as a passive safety features. The preliminary safety analysis of HTR50S for the rupture of co-axial hot gas duct in primary cooling system and the tube rupture of steam generator was conducted to confirm the adequacy of the safety design. It was confirmed that the analysis results satisfied the acceptance criteria.
Pitcher, C. S.*; Barnsley, R.*; Bertalot, L.*; Encheva, A.*; Feder, R.*; Friconneau, J. P.*; Hu, Q.*; Levesy, B.*; Loesser, G. D.*; Lyublin, B.*; et al.
Fusion Science and Technology, 64(2), p.118 - 125, 2013/08
Times Cited Count：4 Percentile：32.48(Nuclear Science & Technology)The port-based plasma diagnostic infrastructure on ITER is described, including the port plugs, the interspace support structure and port cell structure. These systems are modular in nature with standardized dimensions. The design of the equatorial and upper port plugs and their modules is discussed, as well as the dominant loading mechanisms. The port infrastructure design has now matured to the point that port plugs are now being populated with multiple diagnostics supplied by a number of ITER partners - two port plug examples are given.
Pitcher, C. S.*; Barnsley, R.*; Feder, R.*; Hu, Q.*; Loesser, G. D.*; Lyublin, B.*; Padasalagi, S.*; Pak, S.*; Reichle, R.*; Sato, Kazuyoshi; et al.
Fusion Engineering and Design, 87(5-6), p.667 - 674, 2012/08
Times Cited Count：12 Percentile：66.54(Nuclear Science & Technology)Sato, Kazuyoshi; Yaguchi, Eiji; Pitcher, C. S.*; Walker, C.*; Encheva, A.*; Kawano, Yasunori; Kusama, Yoshinori
Fusion Engineering and Design, 86(6-8), p.1264 - 1267, 2011/10
Times Cited Count：3 Percentile：26.02(Nuclear Science & Technology)no abstracts in English
Nakasone, Yuji*; Sato, Kazuyoshi; Takahashi, Yukio*
Proceedings of 2010 ASME Pressure Vessels and Piping Conference (PVP 2010) (CD-ROM), 5 Pages, 2010/07
no abstracts in English
Sato, Kazuyoshi; Omori, Junji; Kondoh, Takashi; Hatae, Takaki; Kajita, Shin*; Ishikawa, Masao; Neyatani, Yuzuru; Ebisawa, Katsuyuki*; Kusama, Yoshinori
Fusion Engineering and Design, 84(7-11), p.1713 - 1715, 2009/06
Times Cited Count：1 Percentile：10.22(Nuclear Science & Technology)Engineering analyses have been performed for the representative diagnostic upper port plug of ITER. Maintenance and integration design have been also carried out for the diagnostic components to be installed in the upper port plug. From the electromagnetic and structural analyses, it has come up an important problem to suppress the displacement of the upper port plug rather than to reduce the produced stress. Reducing the EM force will help to decrease the severity of potential displacement. Maximum displacement of the port plug decreases with increasing in the number of slits in a manner that the displacement would seem to be less than the design tolerance. A proposed low body roller and inner frame may enhance maintenance and integration. These studies and designs have established the design basis for the diagnostic upper port plug.
Sato, Kazuyoshi; Omori, Junji; Ebisawa, Katsuyuki*; Kusama, Yoshinori; Neyatani, Yuzuru
Plasma and Fusion Research (Internet), 2, p.S1088_1 - S1088_4, 2007/11
A part of diagnostic device in vacuum vessel is planned to install in the port plug to make sure the line of sight of diagnostics. Only basic concept is shown for the port plug since design of diagnostic devices has not been substantiated yet. The integration design of the port plug has been performed and the structure concept for electro-magnetic and neutron load has been investigated as for the No.11 upper port plug to confirm reliability of the proposed design. Three diagnostics will be installed in the No.11 upper port plug, the edge Thomson scattering system, the visible-IR TV divertor viewing system and the neutron activation system. To integrate theses diagnostic systems in the port plug, it was designed the arrangement of the labyrinth of optical path, the driving mechanism and cooling systems for shutters and mirrors, the maintenance space. The part just behind the blanket shield module (BSM) was changed to secure a space for maintenance and for associated diagnostic first mirror and shutter, whereas this place is assigned for neutron shielding in the present design. The BSM support, which is main component to apply the electro-magnetic load, was arranged with optical path inside BSM.
Hatae, Takaki; Nakatsuka, Masahiro*; Yoshida, Hidetsugu*; Ebisawa, Katsuyuki*; Kusama, Yoshinori; Sato, Kazuyoshi; Katsunuma, Atsushi*; Kubomura, Hiroyuki*; Shinobu, Katsuya*
Fusion Science and Technology, 51(2T), p.58 - 61, 2007/02
Times Cited Count：6 Percentile：42.5(Nuclear Science & Technology)no abstracts in English
Sato, Kazuyoshi; Uehara, Masaharu*; Tamura, Kosaku*; Hashimoto, Masayoshi*; Ogino, Shunji*; Yagenji, Akira; Nagamatsu, Nobuhide*; Sekiya, Shigeki*; Takahashi, Hideo*; Motohashi, Keiichi*; et al.
JAEA-Technology 2006-024, 114 Pages, 2006/03
JAEA-Technology-2006-024.pdf:24.72MB
no abstracts in English
Sato, Kazuyoshi; Hashimoto, Masayoshi*; Nagamatsu, Nobuhide*; Yagenji, Akira; Sekiya, Shigeki*; Takahashi, Hideo*; Motohashi, Keiichi*; Ogino, Shunji*; Kataoka, Takahiro*; Ohashi, Hironori*; et al.
JAEA-Technology 2006-006, 587 Pages, 2006/03
JAEA-Technology-2006-006.pdf:46.04MB
no abstracts in English
Suzuki, Satoshi; Ezato, Koichiro; Hirose, Takanori; Sato, Kazuyoshi; Yoshida, Hajime; Enoeda, Mikio; Akiba, Masato
Fusion Engineering and Design, 81(1-7), p.93 - 103, 2006/02
Times Cited Count：12 Percentile：63.1(Nuclear Science & Technology)This paper presents an R&D activity on the plasma facing components (PFCs), such as first wall and divertor, for the fusion power plant. The PFCs of the power plant will be subjected to heavy neutron irradiation and high heat/particle flux from plasma during the continuous operation. In the present design of the PFCs, the candidate structural material is a reduced activation ferritic-martensitic steel, F82H, from the viewpoints of low activation and high robustness against neutron irradiation, and the candidate armor material is tungsten from the low sputtering yield and low tritium retention points of view. To realize the PFCs using such materials, JAERI has bee extensively conducting R&Ds on; (1) high performance cooling tube, (2) tungsten armor materials, (3) selection of a bonding technique for F82H and tungsten materials and (4) evaluation of structural integrity. Recent achievements on these R&Ds are presented.
Sato, Kazuyoshi; Ezato, Koichiro; Taniguchi, Masaki; Suzuki, Satoshi; Akiba, Masato
Journal of Nuclear Science and Technology, 42(7), p.643 - 650, 2005/07
Times Cited Count：4 Percentile：30.51(Nuclear Science & Technology)no abstracts in English
Ezato, Koichiro; Dairaku, Masayuki; Taniguchi, Masaki; Sato, Kazuyoshi; Suzuki, Satoshi; Akiba, Masato; Ibbott, C.*; Tivey, R.*
Fusion Science and Technology, 46(4), p.521 - 529, 2004/12
Times Cited Count：8 Percentile：48.81(Nuclear Science & Technology)Thermal hydraulic tests measuring critical heat flux CHF and pressure drop of an annular tube with twisted fin, "annular swirl tube", have been. This tube consists of two concentric tubes, the outer tube and the inner tube with a twisted fin on its outer surface. Cooling water flows inside of the inner tube first, and then returns into an annulus with a swirl flow at an end-return of the cooling tube. The CHF testing shows the no degradation of CHF of the annular swirl tube in comparison with the conventional swirl tube. A minimum axial velocity of 7.1m/sec is required for 28MW/m
, the ITER design value. Applicability of the JAERI's correlation for the heat transfer to the annular swirl tube is also demonstrated by the comparison of the experimental results with those of the numerical analyses. The friction factor correlation for the annular flow with the twisted fins is made for the hydraulic designing of the vertical target. The least pressure drop at the end-return is obtained by using the hemispherical end-plug. Its radius is the same as that of ID of the outer cooling tube.
Ezato, Koichiro; Dairaku, Masayuki; Taniguchi, Masaki; Sato, Kazuyoshi; Suzuki, Satoshi; Akiba, Masato; Ibbott, C.*; Tivey, R.*
Fusion Science and Technology, 46(4), p.530 - 540, 2004/12
Times Cited Count：14 Percentile：66.09(Nuclear Science & Technology)The first fabrication and heating test of a large-scale CFC monoblock divertor mock-up using annular flow concept have been performed to demonstrate its manufacturability and thermo-mechanical performance. Prior to the fabrication of the mock-up, brazed joint tests between the CFC monoblock and the CuCrZr tube have been carried out to find the suitable heat treatment mitigating loss of the high mechanical strength of the CuCrZr material. Basic mechanical examination on CuCrZr undergoing the brazing heat treatment and FEM analyses are also performed to support the design of the mock-up. High heat flux tests on the large-scale divertor mock-up have been performed in an ion beam facility. The mock-up has successfully withstood more than 1,000 thermal cycles of 
for 15 s and 3,000 cycles more than 
for 15 s, which simulates the heat load condition of the ITER divertor. No degradation of the thermal performance of the mock-up has been observed throughout the thermal cycle test.
Ezato, Koichiro; Suzuki, Satoshi; Sato, Kazuyoshi; Akiba, Masato
Journal of Nuclear Materials, 329-333(1), p.820 - 824, 2004/08
Times Cited Count：6 Percentile：40.72(Materials Science, Multidisciplinary)This paper presents thermal fatigue experiments of a cooling tube with a helical triangular fin on its inner cooled surface, namely a 
. The screw thread is directly shaped in a CuCrZr heat sink bar as a cooling channel. Slits with the width of 1.5 mm are machined at the heated side of the heat sink. The thermal fatigue experiments are carried out at 20 and 30 
by using an electron beam irradiation facility in JAERI. Water leakages from fatigue cracks, which locate at the slit of the heat sink, occurred at around 4500th and 1400th cycles at 20 and 30 , respectively. These results show good agreement with lifetime predictions using Manson-Coffin's law based on finite element analyses. Fractographic observations reveal that the fatigue cracks start from the outer heated surface at the slit region of the cooling channel and propagate toward its inner surface.
