JSME Series in Thermal and Nuclear Power Generation Vol. 3; Sodium-cooled fast reactor development; Joyo, Monju, and demonstration reactor

Ohno, Shuji; Maeda, Seiichiro

Dai-27-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 3 Pages, 2023/09

Annual report of Engineering Services Department on JFY2019

Engineering Services Department

JAEA-Review 2021-011, 86 Pages, 2021/08

JAEA-Review-2021-011.pdf:5.35MB

The Engineering Services Department is in charge of operation and maintenance of utility facilities (water distribution systems, electricity supply systems, steam generation systems and drain water systems etc.) in whole of the institute. And also is in charge of operation and maintenance of specific systems (power receiving and transforming facilities, an emergency electric power supply system, an air/liquid waste treatment system, a compressed air supply system) in nuclear reactor facilities, nuclear fuel treatment facilities and usual facilities or buildings. In addition, the department is in charge of maintenance of buildings, design and repair of electrical/mechanical equipments. This annual report describes summary of activities, operation and maintenance data and technical developments of the department carried out in JFY 2019. We hope that this report may help to future work.

Annual report of Department of Research Reactor and Tandem Accelerator, JFY2017 (Operation, utilization and technical development of JRR-3, JRR-4, NSRR, Tandem Accelerator, RI Production Facility and Tritium Process Laboratory)

Department of Research Reactor and Tandem Accelerator

JAEA-Review 2020-073, 113 Pages, 2021/03

JAEA-Review-2020-073.pdf:3.87MB

The Department of Research Reactor and Tandem Accelerator is in charge of the operation, utilization and technical development of JRR-3 (Japan Research Reactor No.3), JRR-4 (Japan Research Reactor No.4), NSRR (Nuclear Safety Research Reactor), Tandem Accelerator, RI Production Facility and Tritium Process Laboratory. This annual report describes the activities of our department in fiscal year of 2017. We carried out the operation and maintenance, utilization, upgrading of utilization techniques, safety administration and international cooperation. Also contained are lists of publications, meetings, granted permissions on laws and regulations concerning atomic energy, outcomes in service and technical developments and so on.

High temperature gas-cooled reactors

Takeda, Tetsuaki*; Inagaki, Yoshiyuki; Aihara, Jun; Aoki, Takeshi; Fujiwara, Yusuke; Fukaya, Yuji; Goto, Minoru; Ho, H. Q.; Iigaki, Kazuhiko; Imai, Yoshiyuki; et al.

High Temperature Gas-Cooled Reactors; JSME Series in Thermal and Nuclear Power Generation, Vol.5, 464 Pages, 2021/02

https://doi.org/10.1016/C2018-0-05383-9

As a general overview of the research and development of a High Temperature Gas-cooled Reactor (HTGR) in JAEA, this book describes the achievements by the High Temperature Engineering Test Reactor (HTTR) on the designs, key component technologies such as fuel, reactor internals, high temperature components, etc., and operational experience such as rise-to-power tests, high temperature operation at 950C, safety demonstration tests, etc. In addition, based on the knowledge of the HTTR, the development of designs and component technologies such as high performance fuel, helium gas turbine and hydrogen production by IS process for commercial HTGRs are described. These results are very useful for the future development of HTGRs. This book is published as one of a series of technical books on fossil fuel and nuclear energy systems by the Power Energy Systems Division of the Japan Society of Mechanical Engineers.

Excellent feature of Japanese HTGR technologies

Nishihara, Tetsuo; Yan, X.; Tachibana, Yukio; Shibata, Taiju; Ohashi, Hirofumi; Kubo, Shinji; Inaba, Yoshitomo; Nakagawa, Shigeaki; Goto, Minoru; Ueta, Shohei; et al.

JAEA-Technology 2018-004, 182 Pages, 2018/07

JAEA-Technology-2018-004.pdf:18.14MB

Research and development on High Temperature Gas-cooled Reactor (HTGR) in Japan started since late 1960s. Japan Atomic Energy Agency (JAEA) in cooperation with Japanese industries has researched and developed system design, fuel, graphite, metallic material, reactor engineering, high temperature components, high temperature irradiation and post irradiation test of fuel and graphite, high temperature heat application and so on. Construction of the first Japanese HTGR, High Temperature engineering Test Reactor (HTTR), started in 1990. HTTR achieved first criticality in 1998. After that, various test operations have been carried out to establish the Japanese HTGR technologies and to verify the inherent safety features of HTGR. This report presents several system design of HTGR, the world-highest-level Japanese HTGR technologies, JAEA's knowledge obtained from construction, operation and management of HTTR and heat application technologies for HTGR.

Activities of the safety and operation project for the international research and development of the sodium-cooled fast reactor in the Generation IV international forum

Sakai, Takaaki; Ren, L.*; Tsige-Tamirat, H.*; Vasile, A.*; Kang, S.-H.*; Ashurko, Y.*; Fanning, T.*

Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), 7 Pages, 2016/06

https://doi.org/10.1115/ICONE24-60502

Annual report of Department of Research Reactor and Tandem Accelerator, JFY2013; Operation, utilization and technical development of JRR-3, JRR-4, NSRR, Tandem Accelerator and RI Production Facility

Department of Research Reactor and Tandem Accelerator

JAEA-Review 2014-047, 153 Pages, 2015/02

JAEA-Review-2014-047.pdf:23.43MB

The Department of Research Reactor and Tandem Accelerator is in charge of the operation, utilization and technical development of JRR-3, JRR-4, NSRR, Tandem Accelerator and RI Production Facility. This annual report describes a summary of activities of services and technical developments carried out in the period between April 1, 2013 and March 31, 2014.

Analytical results of coolant flow reduction test in the HTTR

Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Iyoku, Tatsuo

Proceedings of 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-11) (CD-ROM), 12 Pages, 2005/10

Safety demonstration tests using the HTTR are in progress to verify the inherent safety features, to improve the safety design and the technologies for High Temperature Gas-cooled Reactors (HTGRs). The coolant flow reduction test by tripping one or two out of three gas circulators is one of the safety demonstration tests. The reactor power safely becomes a stable level without a reactor scram and the temperature transient of the reactor-core is very slow. The SIRIUS code was developed to analyze reactor transient during the tests with reactor dynamics. This paper describes the validation of the SIRIUS code with the measured values of one and two gas circulators tripping test at 30% (9 MW). It was confirmed that the SIRIUS code was able to analyze the reactor transient within 10% during the tests. The result of this study and the way of resolving problems can be applied to development for not only the commercial HTGRs but also the Very High Temperature Reactor (VHTR) as one of the Generation IV reactors.

Proposal for evaluation methods of reactor outlet coolant temperature in HTGRs

Takamatsu, Kuniyoshi; Nakagawa, Shigeaki

JAERI-Tech 2005-030, 21 Pages, 2005/05

JAERI-Tech-2005-030.pdf:1.06MB

The High Temperature engineering Test Reactor (HTTR) is a graphite moderated and gas cooled reactor with the thermal power of 30MW and the reactor outlet coolant temperature of 850C/950C. Rise-to-power test in the HTTR was performed from March 31th to May 1st in 2004 as phase 5 test up to 30MW in the high temperature test operation mode. It was confirmed that the thermal reactor power and the reactor outlet coolant temperature reached to 30MW and 950C respectively on April 19th. Achievement of the reactor outlet coolant temperature of 950C is the first time in Japan as well as the world. This report describes proposal for evaluation methods of reactor outlet coolant temperature in the HTGRs through the HTTR operation experiments. The equation is derived from relationships among PRM reading values, reactor outlet coolant temperature, reactor thermal power and heat removal by VCS. The deliberation processes in this study will be applicable to the research and developments of HTGRs in the future.

Radiation monitoring data of the HTTR rise-to-power test; Results up to 30MW operation on the high-temperature test operation mode

Ashikagaya, Yoshinobu; Kawasaki, Tomokatsu; Yoshino, Toshiaki; Ishida, Keiichi

JAERI-Tech 2005-010, 81 Pages, 2005/03

JAERI-Tech-2005-010.pdf:16.65MB

no abstracts in English