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JAEA Reports

Prototype fast breeder reactor Monju; Its history and achievements

Tsuruga Comprehensive Research and Development Center

JAEA-Technology 2019-007, 159 Pages, 2019/07

JAEA-Technology-2019-007.pdf:19.09MB
JAEA-Technology-2019-007-high-resolution1.pdf:42.36MB
JAEA-Technology-2019-007-high-resolution2.pdf:33.56MB
JAEA-Technology-2019-007-high-resolution3.pdf:38.14MB
JAEA-Technology-2019-007-high-resolution4.pdf:48.82MB
JAEA-Technology-2019-007-high-resolution5.pdf:37.61MB

This report summarizes the history and achievements of the prototype fast breeder reactor Monju. The development of Monju started in 1968 as a prototype reactor following the experimental fast reactor Joyo. The development covers all the activity related to the fast reactor; plant design, mockup tests, construction, operation, and plant management. This report summarizes the history and achievements for 11 technical areas: history and principal achievements, design and construction, operation test, plant safety, core physics, fuel, plant system, sodium technology, materials and mechanical design, plant management, and trouble management.

JAEA Reports

Development of the Unified Cross-section Set ADJ2017

Yokoyama, Kenji; Sugino, Kazuteru; Ishikawa, Makoto; Maruyama, Shuhei; Nagaya, Yasunobu; Numata, Kazuyuki*; Jin, Tomoyuki*

JAEA-Research 2018-011, 556 Pages, 2019/03

JAEA-Research-2018-011.pdf:19.53MB
JAEA-Research-2018-011-appendix1(DVD-ROM).zip:433.07MB
JAEA-Research-2018-011-appendix2(DVD-ROM).zip:580.12MB
JAEA-Research-2018-011-appendix3(DVD-ROM).zip:9.17MB

We have developed a new unified cross-section set ADJ2017, which is an improved version of the unified cross-section set ADJ2010 for fast reactors. The unified cross-section set is used for reflecting information of C/E values (analysis / experiment values) obtained by integral experiment analyses, which are stored in the standard database for FBR core design via the cross-section adjustment methodology, which integrates with the information such as uncertainty (covariance) of nuclear data, uncertainty of integral experiment / analysis, sensitivity of integral experiment with respect to nuclear data. The ADJ2017 is based on Japan's latest nuclear data library JENDL-4.0 as in the previous version of ADJ2010, and it incorporates more information on integral experimental data related to minor actinides (MAs) and degraded plutonium (Pu). In the deveropment of ADJ2010, a total of 643 integral experimental data were analyzed and evaluated, and 488 of integral experimental data were finally selected to be used for the cross-section adjustment. In contrast, we have evaluated a total of 719 anlysis results, and eventually adopted 620 integral experimental data to create ADJ2017. ADJ2017 shows almost the same performance as ADJ2010 for the main neutronic characteristics of conventional sodium-cooled MOX-fuel fast reactors. In addition, for the neutrnic characteristics related to MA and degraded Pu, ADJ2017 improves the C/E values of the integral experimental data, and reduces the uncertainty induced by the nuclear data. ADJ2017 is expected to be widely used in the analysis and design research of fast reactors. Moreover, it is expected that the integral experimental data used for ADJ2017 can be utilized as a standard database of FBR core core design.

Journal Articles

Development of safety design guideline of structures, systems and components of Generation-IV Sodium-cooled Fast Reactor

Okano, Yasushi

Nippon Genshiryoku Gakkai-Shi, 60(12), p.764 - 769, 2018/12

JAEA has developed the second safety design guidelines report, "Safety Design Guidelines on Structures, Systems and Components" for Generation-IV SFR system, following the previously published SFR Safety Design Criteria and the first SFR Safety Design Guidelines report and with the reviews by a technical committee under Atomic Energy Society of Japan. This article explains about 14 key points on reactor core system, coolant system, and containment system and also current consistency of international SFR designs to the safety design guidelines.

Journal Articles

Visiting Professor's Research Division

Nakajima, Norihiro; Aoki, Keiko*

Tokyo Daigaku Jinkobutsu Kogaku Kenkyu Senta 2017-Nendo Kenkyu Nempo, p.51 - 53, 84, 2018/12

Visiting professors research division in the Research into Artifacts, Center for Engineering (RACE) has been conducting research collaboration in Socio-Artifactology and Human-Artifactology, in order to establish the methodology of the fusion research in sociology and science for artifacts engineering for the third era activity of RACE. The division decided to observe how the methodology works in applications with social experiments and numerical experiments for 2017.

Journal Articles

Design of LBE spallation target for ADS Target Test Facility (TEF-T) in J-PARC

Saito, Shigeru; Obayashi, Hironari; Wan, T.; Okubo, Nariaki; Sugawara, Takanori; Endo, Shinya; Sasa, Toshinobu

Proceedings of 13th International Topical Meeting on Nuclear Applications of Accelerators (AccApp '17) (Internet), p.448 - 457, 2018/05

JAEA proposes transmutation of minor actinides by accelerator-driven systems (ADS). To obtain the data for ADS design, JAEA plans to construct the ADS Target Test Facility (TEF-T) within the framework of the J-PARC project. In TEF-T, a 250 kW proton-beam-driven LBE (Lead-Bismuth Eutectic) spallation target will be installed to prepare an irradiation database for candidate ADS structural materials. Design activities of the LBE target and target trolley have been progressed. Conceptual design of hot-cells for LBE target loop maintenance and PIE (Post Irradiation Examination) of irradiated samples have been finished. Two LBE loops were manufactured. One is a loop for TEF-T target mock-up and the other is that for collection of material corrosion characteristics in flowing LBE. Oxygen potential control systems for LBE flow have been also developed. Remote handling tests for the target exchange are underway. In this paper, current activities and studies to realize TEF-T will be presented.

Journal Articles

Study of the calculation method for the elastic follow-up coefficient by inelastic analysis

Watanabe, Sota*; Kubo, Koji*; Okajima, Satoshi; Wakai, Takashi

Nippon Kikai Gakkai M&M 2017 Zairyo Rikigaku Kanfarensu Koen Rombunshu (Internet), p.581 - 585, 2017/10

no abstracts in English

Journal Articles

Challenges for management of radioactively contaminated wastes and volume reduction and reuse/recycling of removed soil derived from the activities for environmental remediation after the Fukushima Daiichi Nuclear Power Station accident, 5; Cost evaluation method for the disposal of low level radioactive waste

Nakata, Hisakazu; Sakai, Akihiro; Amazawa, Hiroya; Sakamoto, Yoshiaki

Nippon Genshiryoku Gakkai-Shi, 59(8), p.447 - 449, 2017/08

Removed soil except those that may be reusable/recyclable would be finally disposed of. A general view is obtained in regards to a disposal concept of low level radioactive wastes generated from research, industrial and medical facilities, for the purpose of contributing to designing final disposal facilities of removed soil. It is analyzed to investigate the issues relating to cost evaluation in order to reasonably carry out that design, referring to a cost evaluation methodology applied to a trench-type disposal facility, which has been planned by JAEA, with impermeable layers.

Journal Articles

Data analysis based upon abduction; For better understanding the result discussion in computational science and engineering

Nakajima, Norihiro

Nippon Genshiryoku Gakkai-Shi, 59(8), p.34 - 38, 2017/08

It is necessary the reading comprehension of output data to utilize the simulation in a design process, besides of the input data preparation. The simulation introduces enormous big data for evaluation. This paper describes data analysis technology in the analysis and the evaluation process of the output. The technology applies the artificial intelligence to minimize the unpredictable issues and oversight. It is based on the artifact engineering, which is a multi-sight abduction methodology, which derives a hypothesis.

Journal Articles

Advanced sodium-cooled fast reactor development regarding GIF safety design criteria

Hayafune, Hiroki; Chikazawa, Yoshitaka; Kamide, Hideki; Iwasaki, Mikinori*; Shoji, Takashi*

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Next Generation Nuclear Systems for Sustainable Development (FR-17) (USB Flash Drive), 11 Pages, 2017/06

Design studies on a next generation sodium-cooled fast reactor (SFR) considering the safety design criteria (SDC) developed in the generation IV international forum (GIF) was summarized. To meet SDC including the lessons learned from the TEPCO's Fukushima Dai-ichi Nuclear Power Plants accident, the heat removal function was enhanced to avoid loss of the function even if any internal events exceeding design basis or severe external event happen. Several design options have been investigated and auxiliary core cooling system using air as ultimate heat sink has been selected as an additional cooling system regarding system reliability and diversification. Even though the next generation SFR already adopts seismic isolation system, main component designs have been improved considering revised earthquake conditions. For other external events, design measures for various external events are taken into account. Reactor building design has been improved and important safety components are diversified and located separately improving independency. Those design studies and evaluations on the next generation sodium-cooled reactor have contributed to the development of safety design guidelines (SDG) which is under discussion in the GIF framework.

Journal Articles

Nuclear thermal design of high temperature gas-cooled reactor with SiC/C mixed matrix fuel compacts

Aihara, Jun; Goto, Minoru; Inaba, Yoshitomo; Ueta, Shohei; Sumita, Junya; Tachibana, Yukio

Proceedings of 8th International Topical Meeting on High Temperature Reactor Technology (HTR 2016) (CD-ROM), p.814 - 822, 2016/11

Japan Atomic Energy Agency (JAEA) has started R&D for apply SiC/C mixed matrix to fuel element of high temperature gas-cooled reactors (HTGRs) to improve oxidation resistance of fuel. Nuclear thermal design of HTGR with SiC/C mixed matrix fuel compacts was carried out as a part of above R&Ds. Nuclear thermal design was carried out based on a small sized HTGR for developing countries, HTR50S. Maximum enrichment of uranium is set to be 10 wt%, because coated fuel particles with 10 wt% uranium have been fabricated in Japan. Numbers of kinds of enrichment and burnable poisons (BPs) were set to be same as those of original HTR50S (3 and 2, respectively). We succeeded in nuclear thermal design of a small sized HTGR which performance was equivalent to original HTR50S, with SiC/C mixed matrix fuel compacts. Based on nuclear thermal design, intactness of coated fuel particles was evaluated to be kept on internal pressure during normal operation.

JAEA Reports

Investigation plan for the Mizunami Underground Research Laboratory Project; Investigation plan for the third medium to long-term research phase

Hama, Katsuhiro; Iwatsuki, Teruki; Matsui, Hiroya; Mikake, Shinichiro; Sasao, Eiji; Osawa, Hideaki

JAEA-Review 2016-004, 38 Pages, 2016/06

JAEA-Review-2016-004.pdf:7.07MB

The Mizunami Underground Research Laboratory project is being pursued by the Japan Atomic Energy Agency to enhance the reliability of relevant disposal technologies through investigations of the deep geological environment within the host crystalline rock at Mizunami City in Gifu Prefecture, central Japan. The present report summarizes the research and development activities planned mainly in the -500m gallery.

JAEA Reports

Preliminary assessment of geological disposal system for spent fuel in Japan; First progress report on direct disposal

Radioactive Waste Processing and Disposal Research Department

JAEA-Research 2015-016, 327 Pages, 2015/12

JAEA-Research-2015-016.pdf:41.98MB

The Japan Atomic Energy Agency has prepared the technical progress report on preliminary assessment of geological disposal for spent fuel (hereinafter referred to as "First Progress Report on Direct Disposal"). This report is aiming to examine the technical feasibility of the direct disposal of spent fuel in Japan, based on the results of research and development (R&D) on SF direct disposal carried out during FY 2013. In the First Progress Report on Direct Disposal, the available technology for the direct disposal of spent fuel in Japan was discussed through the preliminary design and safety assessment for the geological disposal system which were made under the limited conditions of representative characteristics of geological environment and spent fuel. Through R&D, the challenges and concerns on the engineering technology and the safety assessment, to be resolved for the Second Progress Report on Direct Disposal, were identified and classified.

JAEA Reports

Synthesized research report in the second mid-term research phase; Mizunami Underground Research Laboratory Project, Horonobe Underground Research Laboratory Project and Geo-stability Project

Hama, Katsuhiro; Mizuno, Takashi; Sasao, Eiji; Iwatsuki, Teruki; Saegusa, Hiromitsu; Sato, Toshinori; Fujita, Tomoo; Sasamoto, Hiroshi; Matsuoka, Toshiyuki; Yokota, Hideharu; et al.

JAEA-Research 2015-007, 269 Pages, 2015/08

JAEA-Research-2015-007.pdf:68.65MB
JAEA-Research-2015-007(errata).pdf:0.07MB

We have synthesised the research results from Mizunami/Horonobe URLs and geo-stability projects in the second mid-term research phase. It could be used as technical bases for NUMO/Regulator in each decision point from sitting to beginning of disposal (Principal Investigation to Detailed Investigation Phase). High quality construction techniques and field investigation methods have been developed and implemented and these will be directly applicable to the National Disposal Program (along with general assessments of hazardous natural events and processes). It will be crucial to acquire technical knowledge on decisions of partial backfilling and final closure by actual field experiments in Mizunami/Horonobe URLs as main themes for the next phases.

Journal Articles

Safety improvement in building layout design to meet the safety design criteria for the Generation IV SFR

Kato, Atsushi; Chikazawa, Yoshitaka; Nabeshima, Kunihiko; Iwasaki, Mikinori*; Akiyama, Yo*; Oya, Takeaki*

Proceedings of 2015 International Congress on Advances in Nuclear Power Plants (ICAPP 2015) (CD-ROM), p.593 - 600, 2015/05

Japan sodium cooled fast reactor is the advanced loop type reactor developing in Japan. After the Fukushima-Dai-ichi NPP accident, system enhancement against severe accident have been investigated mainly for residual decay heat removal system, spent fuel storage system and emergency power sources in order to satisfy the safety design criteria for Generation IV SFR. This paper describes principle of the building layout design and the actual approach to be consistent with the recent design enhancement in JSFR. From the perspective of greater ability to withstand severe events, the principles of the building layout design as the measures against aircraft attack and the consequential fire, and tsunami are introduced in order to avoid local event initiating and simultaneous redundant failure of the safety grade facilities and could achieve lowering risk of the loss of all stuck and maintaining the essential power supply.

Journal Articles

JSFR design progress related to development of safety design criteria for generation IV sodium-cooled fast reactors, 1; Overview

Kamide, Hideki; Ando, Masato*; Ito, Takaya*

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 7 Pages, 2015/05

JAEA, JAPC and MFBR have been conducted design study for the Japan Sodium-cooled Fast Reactor (JSFR), which is a design concept aiming at future commercial use as sustainable electric power source. Since 2011, in order to contribute to the development of safety design criteria (SDC) and safety design guideline (SDG), which include the lesson learned from the TEPCO's Fukushima Dai-ichi Nuclear Power Plants accident, in the frame work of generation IV international forum (GIF), the design study is focusing on the design measures against sever external events such as earthquake and tsunami. At the same time, the design study is going into detail and paying much attention to the maintenance and repair to make surer its feasibility. This paper summarizes the design concept of the demonstration version of JSFR in which progress of design work was incorporated.

Journal Articles

Current status of R&D activities and future plan of Mizunami Underground Research Laboratory

Osawa, Hideaki; Koide, Kaoru; Sasao, Eiji; Iwatsuki, Teruki; Saegusa, Hiromitsu; Hama, Katsuhiro; Sato, Toshinori

Proceedings of 2015 International High-Level Radioactive Waste Management Conference (IHLRWM 2015) (CD-ROM), p.371 - 378, 2015/04

The Mizunami Underground Research Laboratory (MIU) project, launched as a generic underground research laboratory for crystalline rock in 1996, has proceeded in three overlapping phases, "Phase I: Surface-based investigation", "Phase II: Construction" and "Phase III: Operation". Currently, Phase II construction of research drifts in the MIU has been completed to the -500 m level. Phase III research activities have been conducted underground since 2010. The scientific and technical knowledge and know-how acquired in Phases I and II have been released via a web-based report "CoolRep H26". JAEA will continue to promote R&D activities in Phase III at the MIU to build technical confidence.

JAEA Reports

Study on engineering technologies in the Mizunami Underground Research Laboratory (FY 2013); Development of recovery and mitigation technology on excavation damage (Contract research)

Fukaya, Masaaki*; Hata, Koji*; Akiyoshi, Kenji*; Sato, Shin*; Takeda, Yoshinori*; Miura, Norihiko*; Uyama, Masao*; Kaneda, Tsutomu*; Ueda, Tadashi*; Toda, Akiko*; et al.

JAEA-Technology 2014-040, 199 Pages, 2015/03

JAEA-Technology-2014-040.pdf:37.2MB

The researches on engineering technology in the Mizunami Underground Research Laboratory (MIU) project consists of (1) development of design and construction planning technologies, (2) development of construction technology, (3) development of countermeasure technology, (4) development of technology for security, and (5) development of technologies for restoration and/or reduction of the excavation damage. The researches on engineering technology such as verification of the initial design were being conducted by using data measured during construction as a part of the second phase of the MIU plan. Examination about the plug for reflood test in the GL-500m Access/Research Gallery-North as part of the development of technologies for restoration and/or reduction of excavation damage were carried out. Specifically, Literature survey was carried out about the plug, based on the result of literature survey, examination of the design condition, design of the plug and rock stability using numerical simulation, selection of materials for major parts, and grouting for water inflow from between rock and plug, were carried out in this study.

JAEA Reports

Study on nuclear analysis method for high temperature gas-cooled reactor and its nuclear design (Thesis)

Goto, Minoru

JAEA-Review 2014-058, 103 Pages, 2015/03

JAEA-Review-2014-058.pdf:22.36MB

The following issues were investigated using experimental data of HTTR, which is a Japan's HTGR with 30 MW thermal power. (1)Applicability of nuclear data libraries to nuclear analysis for HTGR, (2) Applicability of the improved nuclear analysis method for HTGR, (3) Effectiveness of a rod-type burnable poison on HTGR reactivity control. Using these investigation results, a nuclear design of a small-sized HTGR with 50 MW thermal power (HTR50S) was performed. In the nuclear design of HTR50S, we challenged to decrease the number of the fuel enrichments and to increase the power density compared with HTTR. As a result, the nuclear design was completed successfully by reducing the number of the fuel enrichment to only three from twelve of HTTR and increasing the power density by 1.4 times of HTTR.

Journal Articles

Tunnel support design for anisotropic stress state and comparison between in-situ convergence results and calculated results

Motoshima, Takayuki*; Yabuki, Yoshio*; Minamide, Masashi*; Nago, Makito*; Aoyagi, Kazuhei

Tonneru Kogaku Hokokushu (CD-ROM), 24, p.I_10_1 - I_10_5, 2014/12

Economic tunnel support design for Horonobe underground research laboratory was obtained according to the relationship in the direction of the initial stress and the direction of excavation. The authors compared between the in situ convergence results and calculated results in order to investigate the validity of initial stress measurements. As a result, a positive correlation was observed between the in situ convergence results and calculated results, and the difference between the two was able to be explained by the difference between the assumed deformation coefficient and the measured coefficient. From these results, the measurement results of the initial stress performed in the surface based investigation has been confirmed almost reasonable.

Journal Articles

Operation scenarios and requirements for fuel processing in future fusion reactor facilities; Hydrogen isotope separation as a key process for fuel recycle and safety

Ohira, Shigeru; Yamanishi, Toshihiko; Hayashi, Takumi

Journal of Nuclear Science and Technology, 43(4), p.354 - 360, 2006/04

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

In this paper, expected operation scenarios for ITER and future fusion reactors from a viewpoint of an integrated isotope processing in a future D-T fusion rector are provided with comparisons of requirements for system design attributed to the operation scenarios, safety requirements, etc. Most of the basic requirements for fuel process of a D-T fusion reactor facility common are the same, but the design requirements coming from the individual operation scenarios of ITER and future demo reactors will differ. The system design requirements of the tritium plant taking care of various operations of ITER and a demo reactor are examined and compared. Due to the increase of tritium concentration in the coolant of a demo reactor by tritium permeation in the structural material of the in-vessel components operated at a temperature higher than that of ITER detritiation of coolant will be getting more important. Some important key parameters related to hydrogen isotope processing in future fusion reactors will be discussed.

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