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Takabe, Yugo; Otsuka, Noriaki; Fuyushima, Takumi; Sayato, Natsuki; Inoue, Shuichi; Morita, Hisashi; Jaroszewicz, J.*; Migdal, M.*; Onuma, Yuichi; Tobita, Masahiro*; et al.
JAEA-Technology 2022-040, 45 Pages, 2023/03
JAEA-Technology-2022-040.pdf:6.61MB
Because of the decommission of the Japan Materials Testing Reactor (JMTR), the domestic neutron irradiation facility, which had played a central role in the development of innovative nuclear reactors and the development of technologies to further improve the safety, reliability, and efficiency of light water reactors, was lost. Therefore, it has become difficult to pass on the operation techniques of the irradiation test reactors and irradiation technologies, and to train human resources. In order to cope with these issues, we conducted a study on the implementation of irradiation tests using overseas reactors as neutron irradiation sites as an alternative method. Based on the "Arrangement between the National Centre for Nuclear Research and the Japan Atomic Energy Agency for Cooperation in Research and Development on Testing Reactor," the feasibility of conducting an irradiation test at the MARIA reactor (30 MW) owned by the National Centre for Nuclear Research (NCBJ) using the temperature control system, which is one of the JMTR irradiation technologies, was examined. As a result, it was found that the irradiation test was possible by modifying the ready-made capsule manufactured in accordance with the design and manufacturing standards of the JMTR. After the modification, a penetration test, an insulation continuity test, and an operation test in the range of room temperature to 300
C, which is the operating temperature of the capsule, were conducted and favorable results were obtained. We have completed the preparations prior to transport to the MARIA reactor.
Hamamoto, Shimpei; Sawahata, Hiroaki; Suzuki, Hisashi; Ishii, Toshiaki; Yanagida, Yoshinori
JAEA-Technology 2017-012, 20 Pages, 2017/06
JAEA-Technology-2017-012.pdf:7.9MB
A melt wire was installed at the tip of the control rod in order to measure the temperature of High Temperature engineering Test Reactor (HTTR). After experience with reactor scram from the state of reactor power 100%, the melt wire was taken out from the control rod and appearance has been observed visually. In this study, an exclusive device for taking out the melt wire was prepared. The take-out device functions as expected, and the melt wire was safely and reliably taken out using a remote manipulator. And because the visual observation of the melt wire was clearly carried out, we were successful in developing the control rod temperature measurement technology. It was confirmed that the melt wires with a melting point of 505C or less were melted, and the melt wires with a melting point of 651C or more were not melted. Therefore, it was found that the highest arrival temperature of tip of the control rods where the melt wires are installed reaches within the range of 505 to 651C. And it was found that the control rod temperature at the time of reactor scram does not exceed the using temperature criteria (900C) of Alloy 800H of the control rod sleeve.
Okumura, Susumu; Arakawa, Kazuo; Fukuda, Mitsuhiro; Nakamura, Yoshiteru; Yokota, Wataru; Ishimoto, Takayuki*; Kurashima, Satoshi; Ishibori, Ikuo; Nara, Takayuki; Agematsu, Takashi; et al.
Review of Scientific Instruments, 76(3), p.033301_1 - 033301_6, 2005/03
Times Cited Count：9 Percentile：43.02(Instruments & Instrumentation)A magnetic field drift, gradual decrease of the order of 10
in several tens of hours, was observed with the beam intensity decrease in an operation of an azimuthally-varying-field (AVF) cyclotron. From our experimental results, we show that the temperature increase of the magnet iron by the heat transfer from the excitation coils can induce such change of the magnetic field as to deteriorate the beam quality. The temperature control of the magnet iron was realized by thermal isolation between the main coil and the yoke and by precise control of the cooling water temperature of the trim coils attached to the pole surfaces in order to prevent temperature change of the magnet iron. The magnetic field stability of 
5
10
and the beam intensity stability of 2% have been achieved by this temperature control.
Tachibana, Yukio; Sawahata, Hiroaki; Iyoku, Tatsuo; Nakazawa, Toshio
Nuclear Engineering and Design, 233(1-3), p.89 - 101, 2004/10
Times Cited Count：10 Percentile：55.63(Nuclear Science & Technology)no abstracts in English
Saito, Kenji; Nakagawa, Shigeaki; Hirato, Yoji; Kondo, Makoto; Sawahata, Hiroaki; Tsuchiyama, Masaru*; Ando, Toshio*; Motegi, Toshihiro; Mizushima, Toshihiko; Nakazawa, Toshio
JAERI-Tech 2004-042, 26 Pages, 2004/04
JAERI-Tech-2004-042.pdf:1.16MB
The reactor control system of HTTR is composed of the reactor power control system, the reactor inlet coolant temperature control system, the primary coolant flow rate control system and so on. The reactor control system of HTTR achieves reactor power 30MW, reactor outlet coolant temperature 850C, reactor inlet coolant temperature 395C under the condition that primary coolant flow rate is fixed. In the Rise-to-Power Test, the performance test of the reactor inlet coolant temperature control system was carried out in order to confirm the control capability of this control system. This report shows the test results of performance test. As a result, the control parameters, which can control the reactor inlet coolant temperature stably during the reactor operation, were successfully selected. And it was confirmed that the reactor inlet coolant temperature control system has the capability of controlling the reactor inlet coolant temperature stably against any disturbances on the basis of operational condition of HTTR.
Nagao, Yoshiharu; Miyazawa, Masataka; Komukai, Bunsaku; Fujiki, Kazuo
JAERI-Tech 2003-067, 33 Pages, 2003/07
JAERI-Tech-2003-067.pdf:1.63MB
no abstracts in English
Takada, Eiji*; Fujimoto, Nozomu; Matsuda, Atsuko*; Nakagawa, Shigeaki
JAERI-Tech 2003-040, 23 Pages, 2003/03
JAERI-Tech-2003-040.pdf:1.28MB
In the High Temperature Engineering Test Reactor (HTTR), since the primary circuit is very high at the high temperature test operation, the special alloy Alloy800H is used as the metallic material for cladding tubes and spines of the control rods to endure the temperature of 950 degrees centigrade. The control rod is supposed to be exchanged for the excess use of its temperature limit 900 degrees centigrade according to the strength data of Alloy800H. The scram shutdown by loss of off-site electric power at the high temperature test operation is assumed as an event of the temperature of the control rods to exceed 900 degrees centigrade. In this report, the temperature of the control rods is analyzed by using the measurement data of the rise-to-power test. The result of this analysis it is confirmed that the control rod temperature does not exceed its limitation value even after the most temperature raises event of the loss of off-site electric power at the high temperature test operation.
Sogabe, Toshiaki; Ishihara, Masahiro; Baba, Shinichi; Kojima, Takao; Tachibana, Yukio; Iyoku, Tatsuo; Hoshiya, Taiji; Hiraoka, Toshiharu*; Yamaji, Masatoshi*
JAERI-Research 2002-026, 22 Pages, 2002/11
JAERI-Research-2002-026.pdf:2.41MB
Carbon Fiber Reinforced Carbon-carbon Composites, C/C composites, have been developed and extensively studied their characteristics. C/C composites are considered to be promising materials for the application of a control rod in the next high performance high temperature gas-cooled reactors. In the present paper, details of the development of the candidate C/C composite are described. In the course of the development of the material, especially, feasibility of the production, stableness of the supply and cost are much taken into consideration. As the physical properties of the material, high mechanical strength such as tensile and bending, high fracture strain and fracture toughness and low dimensional change by neutron irradiation have to be met. The developed 2D-C/C composite consists of plain-weave PAN-based carbon fiber cloth and pitch derived matrix. Also, high purification up to the level of nuclear grade was successfully attained in the composite.
Tobita, Masahiro*; Itabashi, Yukio
JAERI-Tech 2002-042, 40 Pages, 2002/03
JAERI-Tech-2002-042.pdf:2.09MB
In relation to aging of light water reactors (LWRs), Irradiation Assisted Stress Corrosion Cracking (IASCC) has been regarded as a significant and urgent issue for reliability of in-core components of LWRs. It is essential for IASCC studies to irradiate test materials under well-controlled of Boiling Water Reactor (BWR) conditions simulating the in-core environment. Therefore, the study for the design of the new water control unit to supply high temperature water into saturated temperature capsules in the Japan Materials Testing Reactor (JMTR) has been carried out. This report summarizes the results of estimation using ORIGEN-2 and QAD-CGGP2 codes of dose equivalent rate on outer surface of the concrete wall of installation room and dose equivalent rate around the ion-exchangers where the highest dose equivalent rate is expected in the unit after the reactor shutdown.
Kanno, Masaru; Kitajima, Toshio; Homma, Kenzo
KAERI/GP-195/2002, p.71 - 75, 2002/00
Recent irradiation studies aiming at clarifying the detailed mechanisms of irradiation damages to the reactor materials require to maintain the specimens at constant temperature regardless of the reactor power level,in order to avoid artificial effects of temperature transient due to reactor power change. In order to deal with this problem, JMTR has adopted feed-foward control to the gas pressure based on the reactor power signal, and developed new temperature control technique in combination with feedback control of heater power.
Fujimoto, Nozomu; Nojiri, Naoki; Takada, Eiji*; Saito, Kenji; Kobayashi, Shoichi; Sawahata, Hiroaki; Kokusen, Shigeru
JAERI-Tech 2000-091, 49 Pages, 2001/03
no abstracts in English
Okumura, Susumu; Kurashima, Satoshi; Ishimoto, Takayuki*; Yokota, Wataru; Arakawa, Kazuo; Fukuda, Mitsuhiro; Nakamura, Yoshiteru; Ishibori, Ikuo; Nara, Takayuki; Agematsu, Takashi; et al.
Proceedings of 13th Symposium on Accelerator Science and Technology, p.283 - 285, 2001/00
no abstracts in English
C/Zircaloy reaction at temperatures from 1,173 to 1,953KNagase, Fumihisa; Uetsuka, Hiroshi;
Journal of Nuclear Science and Technology, 34(4), p.367 - 374, 1997/04
Times Cited Count：4 Percentile：37.18(Nuclear Science & Technology)no abstracts in English
C and stainless steel at high temperaturesNagase, Fumihisa; Uetsuka, Hiroshi;
Journal of Nuclear Materials, 245(1), p.52 - 59, 1997/00
Times Cited Count：46 Percentile：93.51(Materials Science, Multidisciplinary)no abstracts in English
Tachibana, Yukio; Shiozawa, Shusaku; *; *; *
Nucl. Eng. Des., 172(1-2), p.93 - 102, 1997/00
Times Cited Count：8 Percentile：56.29(Nuclear Science & Technology)no abstracts in English
Nagase, Fumihisa; ; Uetsuka, Hiroshi; Furuta, Teruo
JAERI-M 92-179, 31 Pages, 1992/11
no abstracts in English
; ; Aso, Tomokazu; Niimi, Motoji
JAERI-M 92-149, 78 Pages, 1992/10
no abstracts in English
Nagase, Fumihisa; ; Uetsuka, Hiroshi; Furuta, Teruo
JAERI-M 92-001, 28 Pages, 1992/02
no abstracts in English
Murata, Isao; Yamashita, Kiyonobu; Maruyama, So; Fujimoto, Nozomu; Shindo, Ryuichi; Sudo, Yukio
JAERI-M 91-165, 71 Pages, 1991/10
no abstracts in English
Hada, Kazuhiko; Nishiguchi, Isoharu; ; Tsuji, Hirokazu
Nucl. Eng. Des., 132, p.1 - 11, 1991/00
Times Cited Count：24 Percentile：90.25(Nuclear Science & Technology)no abstracts in English
