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Nuclear System Technology Review Committee
JAEA-Review 2024-018, 38 Pages, 2024/06
JAEA-Review-2024-018.pdf:1.46MB
In the R&D activities related to the Ministry of Education, Culture, Sports, Science and Technology's Innovative Nuclear R&D Program, "Development of Integrated Energy System Simulation Method Utilizing Small Modular Reactors for Enhanced System Decarbonization and Resilience," Japan Atomic Energy Agency (JAEA) established the "Nuclear System Technology Review Committee," consisting of experts in the subject areas, to obtain advice on the feasibility of deploying Design-standardized, Factory-built, Site-independent Small Modular Reactors (DFS-SMRs) in Japan and other countries. The Committee met three times during the 2021-2024 project period to discuss proposals for a regulatory framework for the potential commercial deployment of DFS-SMRs in Japan. The starting point for the Committee's discussions was the view that Japan's nuclear regulatory framework, like most other countries with existing commercial nuclear power plants in operation, focuses on large Light Water Reactors. Another consideration was the Committee's view on the basic structure of the regulatory framework, consistent with other regulatory initiatives around the world. Specifically, that the most effective regulatory frameworks need to be less prescriptive, less technology-dependent, and more performance-based. This report focuses on the United States, which has played a leading role in the deployment of SMRs and other advanced reactors, and summarizes the discussions regarding the proposal for a licensing framework for SMRs in Japan, an analysis of the gaps between Japan's current licensing framework and the proposed framework, and specific recommendations for closing the gaps. The Committee is hopeful that the changes to the regulatory framework proposed in this report will become a reality.
Task Force on Maintenance Optimization of Nuclear Facilities
JAEA-Technology 2022-006, 80 Pages, 2022/06
JAEA-Technology-2022-006.pdf:4.24MB
The Task force on maintenance optimization of nuclear facilities was organized in the Nuclear Science Research Institute (NSRI) of Japan Atomic Energy Agency (JAEA) since November 2020, in order to adequately respond to "the New nuclear regulatory inspection system since FY 2020" and to continuously improve the facility maintenance activities. In 2021, the task force has studied (1) optimization of the importance classification on maintenance and inspection of nuclear facilities, and (2) improvement in setting and evaluation of the performance indicators on safety, maintenance and quality management activities, considering "the Graded approach" that is one of the basic methodologies in the new nuclear regulatory inspection system. Each nuclear facility (research reactors, nuclear fuel material usage facilities, others) in the NSRI will steadily improve their respective safety, maintenance and quality management activities, referring the review results suggested by the task force.
Soba, A.*; Prudil, A.*; Zhang, J.*; Dethioux, A.*; Han, Z.*; Dostal, M.*; Matocha, V.*; Marelle, V.*; Lasnel-Payan, J.*; Kulacsy, K.*; et al.
Proceedings of TopFuel 2021 (Internet), 10 Pages, 2021/10
Udagawa, Yutaka; Yamauchi, Akihiro*; Kitano, Koji*; Amaya, Masaki
JAEA-Data/Code 2018-016, 79 Pages, 2019/01
JAEA-Data-Code-2018-016.pdf:2.75MB
FEMAXI-8 is the latest version of the fuel performance code FEMAXI developed by JAEA. A systematic validation work has been achieved against 144 irradiation test cases, after many efforts have been made, in development of new models, improvements in existing models and the code structure, bug-fixes, construction of irradiation-tests database and other infrastructures.
Nagase, Fumihisa; Sakamoto, Kan*; Yamashita, Shinichiro
Corrosion Reviews, 35(3), p.129 - 140, 2017/08
Times Cited Count：13 Percentile：49.95(Electrochemistry)As the lessons learnt from the accident at the Fukushima Daiichi Nuclear Power Station, advanced cladding materials are being developed to enhance accident tolerance comparing with conventional zirconium alloys. The present paper reviews the progress of the development and summarizes subjects to be solved for the enhanced accident-tolerance fuel cladding, focusing on performance degradation under various corrosive environmental conditions that should be considered in designing the LWR fuel.
Nakajima, Norihiro; Nishida, Akemi; Kawakami, Yoshiaki; Suzuki, Yoshio; Sawa, Kazuhiro; Iigaki, Kazuhiko
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 5 Pages, 2015/05
A numerical analysis controlling and managing system is implemented on K, which controls the modelling process and data treating, although the manager only controls a structural analysis by finite element method. The modeling process is described by the list of function ID and its procedures in a data base. The manager executes the process by order in the list for simulation procedures. The manager controls the intention of an analysis by changing the analytical process one to another. Experiments were carried out with static and dynamic analyses.
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.
Takamatsu, Kuniyoshi; Nakagawa, Shigeaki
JAERI-Data/Code 2005-003, 31 Pages, 2005/06
JAERI-Data-Code-2005-003.pdf:4.83MB
Safety demonstration tests using the High Temperature engineering Test Reactor (HTTR) are in progress to verify the inherent safety features for High Temperature Gas-cooled Reactors (HTGRs). The coolant flow reduction test by tripping gas circulators is one of the safety demonstration tests. The reactor power safely brings to a stable level without a reactor scram and the temperature transient of the reactor-core is very slow. The TAC/BLOOST code was developed to analyze reactor and temperature transient during the coolant flow reduction test taking account of reactor dynamics. This paper describes the validation result of the TAC/BLOOST code with the measured values of gas circulators tripping tests at 30 % (9 MW). It was confirmed that the TAC/BLOOST code was able to analyze the reactor transient during the test.
C in HTTRIyoku, Tatsuo; Nakagawa, Shigeaki; Takamatsu, Kuniyoshi
UTNL-R-0446, p.14_1 - 14_9, 2005/03
no abstracts in English
Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Sakaba, Nariaki; Takada, Eiji*; Tochio, Daisuke; Shimakawa, Satoshi; Nojiri, Naoki; Goto, Minoru; Shibata, Taiju; Ueta, Shohei; et al.
JAERI-Tech 2004-063, 61 Pages, 2004/10
JAERI-Tech-2004-063.pdf:3.14MB
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 in the single operation mode using only the primary pressurized water cooler. The parallel loaded operation mode using the intermediate heat exchanger and the primary pressurized water cooler was performed from June 2nd and JAERI (Japan Atomic Energy Research Institute) obtained the certificate of the pre-operation test on June 24th from MEXT (Ministry of Education Culture Sports Science and Technology) after all the pre-operation tests were passed successfully in the high temperature test operation mode. Achievement of the reactor-outlet coolant temperature of 950C is the first time in the world. It is possible to extend highly effective power generation with a high-temperature gas turbine and produce hydrogen from water with a high-temperature. This report describes the results of the high-temperature test operation of the HTTR.
Nakagawa, Shigeaki; Tachibana, Yukio; Takamatsu, Kuniyoshi; Ueta, Shohei; Hanawa, Satoshi
Nuclear Engineering and Design, 233(1-3), p.291 - 300, 2004/10
Times Cited Count：8 Percentile：48.24(Nuclear Science & Technology)The High Temperature Gas-cooled Reactor (HTGR) is particularly attractive due to its capability of producing high temperature helium gas and due to its inherent safety characteristics. The High Temperature Engineering Test Reactor (HTTR), which is the first HTGR in Japan, was successfully constructed at the Oarai Research Establishment of the Japan Atomic Energy Research Institute. The HTTR achieved full power of 30MW at a reactor outlet coolant temperature of about 850C on December 7, 2001 during the "rise-to-power tests". Two kinds of tests were carried out during the "rise-to-power tests". One is commissioning test to get operation permit by the government and another is test to confirm a performance of the reactor, heat exchanger, control system. From the test results of the "rise-to-power tests" up to 30MW, the functionality of the reactor and the cooling system were confirmed, and it was also confirmed that an operation of reactor facility can be performed safely.
Sakaba, Nariaki; Nakagawa, Shigeaki; Takada, Eiji*; Nojiri, Naoki; Shimakawa, Satoshi; Ueta, Shohei; Sawa, Kazuhiro; Fujimoto, Nozomu; Nakazawa, Toshio; Ashikagaya, Yoshinobu; et al.
JAERI-Tech 2003-043, 59 Pages, 2003/03
JAERI-Tech-2003-043.pdf:2.54MB
HTTR plans a high temperature test operation as the fifth step of the rise-to-power tests to achieve a reactor outlet coolant temperature of 950 degrees centigrade in the 2003 fiscal year. Since HTTR is the first HTGR in Japan which uses coated particle fuel as its fuel and helium gas as its coolant, it is necessary that the plan of the high temperature test operation is based on the previous rise-to-power tests with a thermal power of 30 MW and a reactor outlet coolant temperature at 850 degrees centigrade. During the high temperature test operation, reactor characteristics, reactor performances and reactor operations are confirmed for the safety and stability of operations. This report describes the evaluation result of the safety confirmations of the fuel, the control rods and the intermediate heat exchanger for the high temperature test operation. Also, problems which were identified during the previous operations are shown with their solution methods. Additionally, there is a discussion on the contents of the high temperature test operation. As a result of this study, it is shown that the HTTR can safely achieve a thermal power of 30MW with the reactor outlet coolant temperature at 950 degrees centigrade.
Nakagawa, Shigeaki; Fujimoto, Nozomu; Shimakawa, Satoshi; Nojiri, Naoki; Takeda, Takeshi; Saikusa, Akio; Ueta, Shohei; Kojima, Takao; Takada, Eiji*; Saito, Kenji; et al.
JAERI-Tech 2002-069, 87 Pages, 2002/08
JAERI-Tech-2002-069.pdf:10.12MB
Rise-to-power test in the HTTR has been performed from April 23rd to June 6th in 2000 as phase 1 test up to 10MW, from January 29th to March 1st in 2001 as phase 2 test up to 20MW in the rated operation mode and from April 14th to June 8th in 2001 as phase 3 test up to 20MW in the high temperature test operation mode. Phase 4 test to achieve the thermal reactor power of 30MW started from October 23rd in 2001. On December 7th it was confirmed that the thermal reactor power reached to 30MW and the reactor outlet coolant temperature reached to 850C. JAERI obtained the certificate of pre-operation test from MEXT because all the pre-operation tests by MEXT were passed successfully. From the test results of rise-up-power test up to 30MW, the performance of reactor and cooling system were confirmed, and it was confirmed that an operation of reactor facility could be performed safely. Some problems to be solved were found through tests. By means of solving them, the reactor operation with the reactor outlet coolant temperature of 950C will be achievable.
Iigaki, Kazuhiko; Sakaba, Nariaki; Kawaji, Satoshi; Iyoku, Tatsuo
Transactions of 16th International Conference on Structural Mechanics in Reactor Technology (SMiRT-16) (CD-ROM), 7 Pages, 2001/08
no abstracts in English
JAERI-Research 99-016, 84 Pages, 1999/03
JAERI-Research-99-016.pdf:2.97MB
no abstracts in English
Tanaka, Toshiyuki; Okubo, Minoru; Iyoku, Tatsuo; Kunitomi, Kazuhiko; Takeda, Takeshi; Sakaba, Nariaki; Saito, Kenji
Nihon Genshiryoku Gakkai-Shi, 41(6), p.686 - 698, 1999/00
Times Cited Count：4 Percentile：34.52(Nuclear Science & Technology)no abstracts in English
Sakaba, Nariaki; Iigaki, Kazuhiro; Kawaji, Satoshi; Iyoku, Tatsuo
JAERI-Tech 98-013, 152 Pages, 1998/03
no abstracts in English
Fuketa, Toyoshi; Ishijima, Kiyomi; Tanzawa, Sadamitsu; Nakamura, Takehiko; Sasajima, Hideo; Kashima, Yoichi; ;
JAERI-Research 95-005, 53 Pages, 1995/01
JAERI-Research-95-005.pdf:1.96MB
no abstracts in English
Kyoya, Masahiko; ; Kusunoki, Tsuyoshi; ; Takahashi, Teruo*
JAERI-M 94-079, 116 Pages, 1994/06
no abstracts in English
Arigane, Kenji; ; ; Aoyama, Isao; Seguchi, Tadao; Takahashi, Hidetake
JAERI-M 92-078, 22 Pages, 1992/06
no abstracts in English
