Nagase, Fumihisa; Otomo, Takashi; Uetsuka, Hiroshi*
Nuclear Technology, 208(3), p.484 - 493, 2022/03
An Ag-In-Cd control rod alloy was heated in argon or oxygen at 1073-1673 K for 60-3600 s and the release behavior of the elements was examined. Complete liquefaction of the alloy occurred between 1123 and 1173 K, and elemental release was quite limited below the liquefaction temperature. In argon, almost all of the Cd content was released within 3600 s at 1173 K and within 60 s at 1573 K, while the released fractions of Ag and In were 3% and 8%, respectively. In oxygen, the release of Cd, which was quite small at temperatures up to 1573 K, drastically increased to 30-50% at 1673 K for short periods. Releases of Ag and In were also small in oxygen under the examined conditions. Comparison with the experimental data suggests that conventional empirical release models may underestimate the Cd release at lower temperatures just after control rod failure in severe accidents.
Takino, Kazuo; Sugino, Kazuteru; Oki, Shigeo
Annals of Nuclear Energy, 162, p.108454_1 - 108454_7, 2021/11
Ho, H. Q.; Fujimoto, Nozomu*; Hamamoto, Shimpei; Nagasumi, Satoru; Goto, Minoru; Ishitsuka, Etsuo
Nuclear Engineering and Design, 377, p.111161_1 - 111161_9, 2021/06
Udagawa, Yutaka; Fuketa, Toyoshi*
Comprehensive Nuclear Materials, 2nd Edition, Vol.2, p.322 - 338, 2020/08
Nagasumi, Satoru; Matsunaka, Kazuaki*; Fujimoto, Nozomu*; Ishii, Toshiaki; Ishitsuka, Etsuo
JAEA-Technology 2020-003, 13 Pages, 2020/05
The influence of the control rod model on the nuclear characteristics of the HTTR has been evaluated, by creating detailed control rod model, in which geometric shape was close to that of the actual control rod structure, in MVP code. According to refinement of the control rod model, the critical control rod position was 11 mm lower than that of the conventional model, and this was close to the measured value of 1775 mm. The reactivity absorbed by the shock absorber located at the tip of the control rod was 0.2%k/k, and this was 14 mm difference at the critical control rod position. Considering the effect of refinement of the control rod and the effect of the shock absorber, the correction amount for the analysis value in SRAC code due to the shape effect of the control rod, is -0.05%k/k in reactivity, and -3 mm in the critical control rod position at low temperature criticality.
Miwa, Shuhei; Takase, Gaku; Imoto, Jumpei; Nishioka, Shunichiro; Miyahara, Naoya; Osaka, Masahiko
Journal of Nuclear Science and Technology, 57(3), p.291 - 300, 2020/03
For the evaluation of transport behavior of control material boron in a severe accident of BWR from the viewpoint of chemical effects on cesium and iodine behavior, boron chemistry during transportation in the high temperature region above 400 K was experimentally investigated. The heating tests of boron oxide specimen were conducted using the dedicated experimental apparatus reproducing fission product release and transport in steam atmosphere. Released boron oxide vapor was deposited above 1,000 K by the condensation onto stainless steel. The boron deposits and/or vapors significantly reacted with stainless steel above 1,000 K and formed the stable iron-boron mixed oxide (FeO)BO. These results indicate that released boron from degraded BWR control blade in a severe accident could remain in the high temperature region such as a Reactor Pressure Vessel. Based on these results, it can be said that the existence of boron deposits in the high temperature region would decrease the amount of transported cesium vapors from a Reactor Pressure Vessel due to possible formation of low volatile cesium borate compounds by the reaction of boron deposits with cesium vapors.
Kato, Shinya; Matsuba, Kenichi; Kamiyama, Kenji; Ganovichev, D. A.*; Baklanov, V. V.*
Proceedings of 2019 International Congress on Advances in Nuclear Power Plants (ICAPP 2019) (Internet), 9 Pages, 2019/05
In order to ensure In-Vessel Retention (IVR) of molten-core in Core Disruptive Accident (CDA), we are investigating the possibility of the molten-core discharge through the control rod guide tube (CRGT) to prevent energetics due to exceeding the prompt criticality. Internal structures of the CRGT, such as a sodium-flow regulator when the CRGT is connected to the high-pressure plenum, may disturb the discharge of molten-core from the core region. Based on above background, an experimental program to clarify characteristics of molten-core discharge through the CRGT has been commenced as one of subjects under a joint study with National Nuclear Center of the Republic of Kazakhstan (NNC-RK) named EAGLE-3 project. An experiment using molten-alumina as fuel simulant and sodium was conducted at the out-of-pile test facility owned by NNC-RK to investigate sodium cooling effect around the sodium flow regulator on its destruction. The experimental result represented that void development at the initiation of molten-alumina discharge eliminated liquid-phase sodium from the discharge path and this also eliminated sodium cooling effect around the sodium flow regulator. As a result, early destruction of the sodium flow regulator and massive discharge of molten alumina occurred in turn.
Sasaki, Shinji; Maeda, Koji; Furuya, Hirotaka*
Journal of Nuclear Science and Technology, 55(3), p.276 - 282, 2018/03
Hamamoto, Shimpei; Sawahata, Hiroaki; Suzuki, Hisashi; Ishii, Toshiaki; Yanagida, Yoshinori
JAEA-Technology 2017-012, 20 Pages, 2017/06
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.
Nishihara, Tetsuo; Inagaki, Yoshiyuki
Nuclear Technology, 153(1), p.100 - 106, 2006/01
Japan Atomic Energy Research Institute (JAERI) has performed the research and development of hydrogen production using the high temperature engineering test reactor (HTTR). One of the key issues for the HTTR hydrogen production system is the development of control technology for stable operation. A thermal load absorber concept using a steam generator installed downstream of a reformer is proposed to mitigate a variation of helium temperature. Thermal hydraulic analyses for the start up operation and the suspension of feed gas supply to the reformer are carried out. These results show that a large variation of the reformer outlet helium temperature takes place due to a change of the feed gas flow rate. However the steam generator can mitigate the variation of helium temperature. It is clarified that the HTTR can continue normal operation independently of the feed gas flow rate.
JRR-4 Operation Division; Research Reactor Utilization Division
JAERI-Tech 2005-042, 58 Pages, 2005/07
Japan Research Reactor No.4 (JRR-4) was shut down manually, due to the control rod insertion failure occurred during the rated power (3,500kW) operation on June 10, 2005. It became evident by the investigation that a screw bolt at the control rod support got loose and blocked the control rod insertion. The failure was recovered through replacement with the new screw bolt. Considering the importance of this event, we decided to inspect all screw bolts over the core that may cause a control rod insertion failure. Furthermore, we decided to carry out periodical inspection about these screw bolts whether they were tightened enough or not. This report describes the result of inspection carried out as the preventive measures.
Ono, Tomio*; Subekti, M.*; Kudo, Kazuhiko*; Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Nabeshima, Kunihiko
Nihon Genshiryoku Gakkai Wabun Rombunshi, 4(2), p.115 - 126, 2005/06
Control-rod withdrawal tests simulating reactivity insertion are carried out in the HTTR to verify the inherent safety features of HTGRs. This paper describes pre-test analysis method using artificial neural networks to predict the changes of reactor power and reactivity. The network model applied in this study is based on recurrent neural networks. The inputs of the network are the changes of the central control rods position and other significant core parameters, and the outputs are the changes of reactor power and reactivity. Furthermore, Time Synchronizing Signal(TSS) is added to input to improve the modeling of time series data. The actual tests data, which were previously carried out in the HTTR, were used for learning the model of the plant dynamics. After the learning, the network can predict the changes of reactor power and reactivity in the following tests.
Inaba, Yoshitomo; Ohashi, Hirofumi; Nishihara, Tetsuo; Sato, Hiroyuki; Inagaki, Yoshiyuki; Takeda, Tetsuaki; Hayashi, Koji; Takada, Shoji
Nuclear Engineering and Design, 235(1), p.111 - 121, 2005/01
Prior to the connection of a hydrogen production plant to the HTTR, the fluctuation tests of the chemical reaction in the steam reformer with the mock-up test facility of the HTTR hydrogen production system were carried out for the establishment and demonstration of the control technology. As a result, it was shown that the HTTR hydrogen production system with the same control system as the mock-up test facility can provide stable controllability for any disturbance at the steam reformer without the influence to the reactor. In addition, a dynamic simulation code for the HTTR hydrogen production system was verified with the obtained test data.
Tachibana, Yukio; Sawahata, Hiroaki; Iyoku, Tatsuo; Nakazawa, Toshio
Nuclear Engineering and Design, 233(1-3), p.89 - 101, 2004/10
no abstracts in English
Nojiri, Naoki; Shimakawa, Satoshi; Fujimoto, Nozomu; Goto, Minoru
Nuclear Engineering and Design, 233(1-3), p.283 - 290, 2004/10
This paper describes the results of core physics test in start-up and power-up of the HTTR. The tests were conducted in order to ensure performance and safety of the high temperature gas cooled reactor, and was carried out to measure the critical approach, the excess reactivity, the shutdown margin, the control rod worth, the reactivity coefficient, the neutron flux distribution and the power distribution. The expected core performance and the required reactor safety characteristics were verified from the results of measurements and calculations.
Takeda, Takeshi; Tachibana, Yukio
Nuclear Engineering and Design, 223(1), p.25 - 40, 2003/07
no abstracts in English
Takada, Eiji*; Fujimoto, Nozomu; Matsuda, Atsuko*; Nakagawa, Shigeaki
JAERI-Tech 2003-040, 23 Pages, 2003/03
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.
Nishihara, Tetsuo; Inagaki, Yoshiyuki
Proceedings of GLOBAL2003 Atoms for Prosperity; Updating Eisenhower's Global Vision for Nuclear Energy (CD-ROM), p.320 - 324, 2003/00
HTGR hydrogen production system has potential possibility to provide hydrogen without CO emission. Key technology for developing this system is to establish the control technology for preventing propagation of thermal turbulence from the hydrogen production system to the HTGR. Japan Atomic Energy Research Institute (JAERI) has planed a demonstration test of hydrogen production using an HTGR named high temperature engineering test reactor (HTTR) to develop the control technology. Thermal load absorber concept using the steam generator located downstream of the chemical reactor is proposed to mitigate the variation of outlet helium temperature of the chemical reactor. This concept leads to the stable controllability and enables to operate the HTGR and the hydrogen production plant independently. Plant simulation analyses are carried out to verify the performance of this concept.
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
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.
Takeda, Takeshi; Nakagawa, Shigeaki; Homma, Fumitaka*; Takada, Eiji*; Fujimoto, Nozomu
Journal of Nuclear Science and Technology, 39(9), p.986 - 995, 2002/09
no abstracts in English