JOPSS - Search Results

Search Results: Records 1-20 displayed on this page of 78

Presentation/Publication Type

Initialising ...

Refine

Journal/Book Title

Initialising ...

Meeting title

Initialising ...

First Author

Initialising ...

Keyword

Initialising ...

Language

Initialising ...

Publication Year

Initialising ...

Held year of conference

Initialising ...

JAEA Reports

Irradiation test using foreign reactor, 1; Study of irradiation test with capsule temperature control system (Joint research)

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 $^{circ}$ 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.

Journal Articles

Study on the discharge behavior of the molten-core materials through the control rod guide tube; Investigations of the effect of an internal structure in the control rod guide tube on the discharge behavior

Kato, Shinya; Matsuba, Kenichi; Kamiyama, Kenji; Akaev, A.*; Vurim, A.*; Baklanov, V.*

Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-13) (Internet), 12 Pages, 2022/09

The In-Vessel Retention (IVR) of molten-core in Core Disruptive Accidents (CDAs) is of prime importance in enhancing the safety of sodium-cooled fast reactors. One of the main subjects in ensuring IVR is to design the Control Rod Guide Tube (CRGT) which allows effective discharge of molten core materials from the core region. The effectiveness of the CRGT design is assessed through CDA analyses, and it is reasonable for these analyses to develop a computer code collaborated with experimental researches. Thus, experiments addressing the discharge behavior of the molten-core materials through the CRGT have proceeded as one of the subjects in the collaboration research named the EAGLE-3 project, and the obtained experimental results are reflected in the development of the SIMMER code. In this project, a series of out-of-pile tests using molten-alumina as the fuel simulant was conducted to understand the discharge behavior of molten-core materials through the CRGT. In this study, in order to investigate the effect of an internal structure in the CRGT on the discharge behavior of the molten-core materials, the data of an out-of-pile test in which the molten-alumina penetrated to a duct with the internal structure were analyzed. In addition, the post-test analysis using the SIMMER code was conducted and the results were compared with the test results.

Journal Articles

An Investigation on the control rod homogenization method for next-generation fast reactor cores

Takino, Kazuo; Sugino, Kazuteru; Oki, Shigeo

Annals of Nuclear Energy, 162, p.108454_1 - 108454_7, 2021/11

https://doi.org/10.1016/j.anucene.2021.108454

Times Cited Count：1 Percentile：10.83(Nuclear Science & Technology)

Journal Articles

Transient response of LWR fuels (RIA)

Udagawa, Yutaka; Fuketa, Toyoshi*

Comprehensive Nuclear Materials, 2nd Edition, Vol.2, p.322 - 338, 2020/08

https://doi.org/10.1016/B978-0-12-803581-8.11682-0

Journal Articles

Boron chemistry during transportation in the high temperature region of a boiling water reactor under severe accident conditions

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

https://doi.org/10.1080/00223131.2019.1671913

Times Cited Count：9 Percentile：64.25(Nuclear Science & Technology)

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) $_{2}$ BO $_{3}$ . 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.

JAEA Reports

Preventive measures for the recurrence of control rod insertion failure in Japan Research Reactor No.4 (JRR-4)

JRR-4 Operation Division; Research Reactor Utilization Division

JAERI-Tech 2005-042, 58 Pages, 2005/07

JAERI-Tech-2005-042.pdf:5.4MB

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.

Journal Articles

Overview of goals and performance of ITER and strategy for plasma-wall interaction investigation

Shimada, Michiya; Costley, A. E.*; Federici, G.*; Ioki, Kimihiro*; Kukushkin, A. S.*; Mukhovatov, V.*; Polevoi, A. R.*; Sugihara, Masayoshi

Journal of Nuclear Materials, 337-339, p.808 - 815, 2005/03

https://doi.org/10.1016/j.jnucmat.2004.09.063

Times Cited Count：68 Percentile：96.40(Materials Science, Multidisciplinary)

ITER is an experimental fusion reactor for investigation and demonstration of burning plasmas, characterised of its heating dominated by alpha-particle heating. ITER is a major step from present devices and an indispensable step for fusion reactor development. ITER's success largely depends on the control of plasma-wall interactions(PWI), with power and particle fluxes and time scales one or two orders of magnitude larger than in present devices. The strategy for control of PWI includes the semi-closed divertor, strong fuelling and pumping, disruption and ELM control, replaceable plasma-facing materials and stepwise operation.

JAEA Reports

Test results of the reactor inlet coolant temperature control system of HTTR

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 850 $^{circ}$ C, reactor inlet coolant temperature 395 $^{circ}$ C 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.

JAEA Reports

Super safe small reactor RAPID-L conceptual design and R&D, JAERI's nuclear research promotion program, H11-002 (Contract research)

Kobe, Mitsuru*; Tsunoda, Hirokazu*; Mishima, Kaichiro*; Kawasaki, Akira*; Iwamura, Takamichi

JAERI-Tech 2003-016, 68 Pages, 2003/03

JAERI-Tech-2003-016.pdf:4.37MB

The 200 kWe uranium nitride fueled lithium cooled fast reactor "RAPID-L" combined with thermoelectric power conversion system that can be operated unmanned without refueling for up to ten years has been demonstrated. The RAPID refueling concept enables quick and simplified refueling, and achieves plant design lifetime over 20 years. A significant advantage of the RAPID-L design, which does not require the use of control rods - is the introduction of the innovative reactivity control systems: lithium expansion module (LEM), lithium injection module (LIM) and lithium release module (LRM). LEM is the most promisiong candidate for improving inherent reactivity feedback. LEMs could realize burnup compensation. LIMs assure sufficient negative reactivity feedback in unprotected transients. LRMs enable an automated reactor startup by detecting the hot standby temperature of the primary coolant. All these systems use $^{6}$ Li as liquid poison and are actuated by highly reliable physical properties (volume expansion of $^{6}$ Li for LEM, and freeze seal melting for LIM and LRM).

JAEA Reports

Development of the carbon fiber reinforced carbon-carbon composite for High Temperature Gas-cooled Reactors

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.

Journal Articles