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JAEA Reports

Progress report on Nuclear Safety Research Center; JFY 2015 - 2017

Nuclear Safety Research Center, Sector of Nuclear Safety Research and Emergency Preparedness

JAEA-Review 2018-022, 201 Pages, 2019/01

JAEA-Review-2018-022.pdf:20.61MB

Nuclear Safety Research Center (NSRC), Sector of Nuclear Safety Research and Emergency Preparedness, Japan Atomic Energy Agency (JAEA) is conducting technical support to nuclear safety regulation and safety research based on the Mid-Long Term Target determined by Japanese government. This report summarizes the research structure of NSRC and the cooperative research activities with domestic and international organizations as well as the nuclear safety research activities and results in the period from JFY 2015 to 2017 on the nine research fields in NSRC; (1) severe accident analysis, (2) radiation risk analysis, (3) safety of nuclear fuels in light water reactors (LWRs), (4) thermohydraulic behavior under severe accident in LWRs, (5) materials degradation and structural integrity, (6) safety of nuclear fuel cycle facilities, (7) safety management on criticality, (8) safety of radioactive waste management, and (9) nuclear safeguards.

JAEA Reports

A Guide to introducing burnup credit, preliminary version (English translation)

Okuno, Hiroshi; Suyama, Kenya; Ryufuku, Susumu*

JAEA-Review 2017-010, 93 Pages, 2017/06

JAEA-Review-2017-010.pdf:2.47MB

There is an ongoing discussion on the application of burnup credit to the criticality safety controls of facilities that treat spent fuels. With regard to such application of burnup credit in Japan, this document summarizes the current technical status of the prediction of the isotopic composition and criticality of spent fuels, as well as safety evaluation concerns and the current status of legal affairs. This report is an English translation of A Guide to Introducing Burnup Credit, Preliminary Version, originally published in Japanese as JAERI-Tech 2001-055 by the Nuclear Fuel Cycle Facility Safety Research Committee.

Journal Articles

Characterization of the insoluble sludge from the dissolution of irradiated fast breeder reactor fuel

Aihara, Haruka; Arai, Yoichi; Shibata, Atsuhiro; Nomura, Kazunori; Takeuchi, Masayuki

Procedia Chemistry, 21, p.279 - 284, 2016/12

BB2015-3214.pdf:0.31MB

 Times Cited Count:1 Percentile:20.72

JAEA Reports

Solvent extraction and release behavior of ruthenium and europium in fire accident conditions in reprocessing plants (Contract research)

Amano, Yuki; Watanabe, Koji; Masaki, Tomoo; Tashiro, Shinsuke; Abe, Hitoshi

JAEA-Technology 2016-012, 21 Pages, 2016/06

JAEA-Technology-2016-012.pdf:1.81MB

To contribute to safety evaluation of fire accident in fuel reprocessing plants, solvent extraction behavior of ruthenium, which could form volatile species, was investigated. Distribution ratios of ruthenium at fire accident conditions were obtained by extraction experiments with several solvent composition at different temperature as parameters. In order to investigate release behavior of ruthenium and europium at fire accident, release ratios of ruthenium and europium were also obtained by solvent combustion experiments.

JAEA Reports

Investigation of release behavior of volatile ruthenium species from thermal decomposition of ruthenium nitrosylnitrate

Abe, Hitoshi; Masaki, Tomoo; Amano, Yuki; Uchiyama, Gunzo

JAEA-Research 2014-022, 12 Pages, 2014/11

JAEA-Research-2014-022.pdf:1.03MB

To contribute safety evaluation of boiling and drying accident of high active liquid waste (HALW) in fuel reprocessing plant, release behavior of Ru, which was considered as an important nuclide for evaluating public dose from the volatile viewpoint, has been investigated. It has been reported that release of Ru becomes conspicuously after HALW is dried up. In this work, to grasp the release behavior of Ru, release ratio of Ru with thermal decomposition of Ru nitrate, which would be in the dried HALW, was measured and release rate constant of Ru from the nitrate was estimated. It was found that the calculation result of release rate of Ru from the nitrate with rise of temperature by using the constant could well simulate the result acquired from the beaker-scale experiment.

Journal Articles

Experiments on the behavior of americium in pyrochemical process

Hayashi, Hirokazu; Akabori, Mitsuo; Minato, Kazuo

Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 3 Pages, 2005/10

For a basis of the future nuclear cycle, it is very important to understand and control the behavior of TRU (Np, Pu, Am, Cm) in the nuclear fuel cycle. Experimental study of pyrochemical process of fuels containing TRU requires the facility having not only shielding for $$gamma$$-ray and neutron but also ability to keep a high purity inert gas atmosphere; because minor actinide chlorides can easily react with oxygen or water vapor in an atmosphere. The module for TRU high temperature chemistry (TRU-HITEC) had been installed to study the basic properties of TRU in the pyrochemical processes. In the present work, the behavior of $$^{241}$$Am in pyrochemical process was investigated by electrochemical methods.

Journal Articles

Investigation on Innovative Water Reactor for Flexible Fuel Cycle (FLWR), 1; Conceptual design

Uchikawa, Sadao; Okubo, Tsutomu; Kugo, Teruhiko; Akie, Hiroshi; Nakano, Yoshihiro; Onuki, Akira; Iwamura, Takamichi

Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 6 Pages, 2005/10

no abstracts in English

JAEA Reports

Investigation on future nuclear power reactors and fuel cycle systems

Otaki, Kiyoshi*; Tanaka, Yoji*; Katsurai, Kiyomichi*; Aoki, Kazuo*

JAERI-Review 2005-035, 79 Pages, 2005/09

JAERI-Review-2005-035.pdf:4.57MB

In order to collect technical information for the assessment on future nuclear power reactors and fuel cycle systems in Japan, investigation has been made on the characteristics and performance of future reactor options including reduced moderation water reactors (RMWRs) and their fuel cycle systems since the fiscal year 1998. The subjects of investigation are divided into three categories; breeder reactors and their fuel cycle, alternative to sodium-cooled FBR systems,plutonium recycling, spent fuel reprocessing and waste disposal. This report is a summary of the investigation carried out so far.

JAEA Reports

Report on the 8th Workshop on the Innovative Water Reactor for Flexible Fuel Cycle; February 10, 2005, Koku-kaikan, Minato-ku, Tokyo

Kobayashi, Noboru; Okubo, Tsutomu; Uchikawa, Sadao

JAERI-Review 2005-029, 119 Pages, 2005/09

JAERI-Review-2005-029.pdf:11.01MB

The research on Innovative Water Reactor for Flexible fuel cycle (FLWR) has been performed in JAERI for the development of future innovative reactors. The workshop on the FLWRs has been held every year since 1998 aiming at information exchange between JAERI and other organizations. The 8th workshop was held on Feb. 10, 2005 under the joint auspices of JAERI and North Kanto and Kanto-Koetsu branches of Atomic Energy Society of Japan with 75 participants. The workshop began with 3 presentations on FLWRs entitled "Framework and Status of Research and Development on FLWRs", "Long-Term Fuel Cycle Scenarios for Advanced Utilization of Plutonium from LWRs", and "Experiments on Characteristics on Hydrodynamics in Tight-Lattice Core". Then 3 lectures followed: "Development of Evaluation Method for Accuracy in Predicting Neutronics Characteristics of Tight-Lattice Core" by Osaka University, "Development of Cost-Reduced Low-Moderation Spectrum Boiling Water Reactor" by Toshiba Corporation and "Design and Analysis on Super-Critical Water Cooled Power Reactors" by Tokyo University.

JAEA Reports

Proceedings of the International symposium NUCEF 2005; February 9-10, 2005, Techno Community Square RICOTTI, Tokai-mura, Ibaraki-ken, Japan

International Symposium NUCEF 2005 Working Group

JAERI-Conf 2005-007, 359 Pages, 2005/08

JAERI-Conf-2005-007.pdf:24.79MB

Japan Atomic Energy Research Institute (JAERI) held the international symposium NUCEF2005 at Techno Community Square RICOTTI in Tokai-mura on February 9 and 10, 2005. This symposium was co-organized by Japan Nuclear Cycle Development Institute (JNC), and Nuclear Fuel Cycle Safety Research Committee authorized the program. Two hundred thirty-nine participants from 11 countries presented fifty-nine papers, and discussed recent research activities and its outputs on waste disposal safety, fuel cycle facility safety including criticality safety, and separation process development. The presented papers are compiled in the proceedings.

Journal Articles

Study on safety evaluation for nuclear fuel cycle facility under accident conditions

Abe, Hitoshi; Tashiro, Shinsuke; Morita, Yasuji

JAERI-Conf 2005-007, p.199 - 204, 2005/08

Source term data for estimating release behavior of radioactive nuclides is necessary to evaluate synthetic safety of nuclear fuel cycle facility under accident conditions, such as fire and criticality. In JAERI, the data has been obtained by performing some demonstration tests. In this paper, the data for the criticality accident in fuel solution obtained from the TRACY experiment, will be mainly reviewed. At 4.5 h after the transient criticality, the release ratio of the iodine were about 0.2% for re-insertion of transient rod at just after transient criticality and about 0.9% for not re-insertion. Similarly the release coefficient and release ratio for Xe were estimated. It was proved that the release ratio of Xe-141 from the solution was over 90% in case that the inverse period was over about 100 (1/s). Furthermore, outline of the study on the fire accident as future plan will be also mentioned.

Journal Articles

Study on safety evaluation for nuclear fuel cycle facility under fire accident conditions

Tashiro, Shinsuke; Abe, Hitoshi; Morita, Yasuji

JAERI-Conf 2005-007, p.348 - 350, 2005/08

Hot test at Rokkasho Reprocessing plant has been started since last year. In addition, construction of the MOX fuel fabrication facility at Rokkasho site is planning. So, the importance of safety evaluation of the nuclear fuel cycle facility is increasing. Under the fire accident, one of the serious postulated accidents in the nuclear fuel cycle facility, the equipments (glove-box, ventilation system, ventilation filters etc.) for the confinement of the radioactive materials within the facility could be damaged by a large amount of heat and smoke released from the combustion source. Therefore, the fundamental data and models calculating for the amount of heat and smoke released from the combustion source under such accident are important for the safety evaluation of the facility. In JAERI, the study focused on the evaluation of amount of heat and smoke released from the combustion source is planning. In this paper, the outline of experimental apparatus, measurement items and evaluation terms are described.

Journal Articles

Concept of Innovative Water Reactor for Flexible Fuel Cycle (FLWR)

Iwamura, Takamichi; Uchikawa, Sadao; Okubo, Tsutomu; Kugo, Teruhiko; Akie, Hiroshi; Nakatsuka, Toru

Proceedings of 13th International Conference on Nuclear Engineering (ICONE-13) (CD-ROM), 8 Pages, 2005/05

In order to ensure sustainable energy supply in the future based on the matured Light Water Reactor (LWR) and coming LWR-Mixed Oxide (MOX) technologies, a concept of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been investigated in Japan Atomic Energy Research Institute (JAERI). The concept consists of two parts in the chronological sequence. The first part realizes a high conversion type core concept, which is basically intended to keep the smooth technical continuity from current LWR without significant gaps in technical point of view. The second part represents the Reduced-Moderation Water Reactor (RMWR) core concept, which realizes a high conversion ratio over 1.0 being useful for the long-term sustainable energy supply through plutonium multiple recycling based on the well-experienced LWR technologies. The key point is that the two core concepts utilize the compatible and the same size fuel assemblies, and hence, the former concept can proceed to the latter in the same reactor system, based flexibly on the fuel cycle circumstances.

JAEA Reports

Progress of nuclear safety research, 2004

Editorial Committee on Nuclear Safety Research Results

JAERI-Review 2005-009, 151 Pages, 2005/03

JAERI-Review-2005-009.pdf:31.04MB

no abstracts in English

JAEA Reports

Proceedings of 2004 Symposium on Nuclear Data; November 11-12, 2004, JAERI, Tokai, Japan

Tahara, Yoshihisa*; Fukahori, Tokio

JAERI-Conf 2005-003, 254 Pages, 2005/03

JAERI-Conf-2005-003.pdf:32.21MB

The 2004 Symposium on Nuclear Data was held at Tokai Research Establishment, Japan Atomic Energy Research Institute (JAERI), on 11th and 12th of November 2004. Japanese Nuclear Data Committee and Nuclear Data Center, JAERI organized this symposium. In the oral sessions, presented were 19 papers on topics of nuclear data for LWR and nuclear fuel cycle, nuclear data for ADS development, experiences from use of JENDL-3.3 and requests to JENDL-4, recent cross section measurements, nuclear data for life and material sciences, and nuclear data needs and activities in the World. In the poster session, presented were 21 papers concerning experiments, evaluations, benchmark tests, applications and so on. Those presented papers are compiled in the proceedings.

JAEA Reports

The Possible role of reduced-moderation water reactors and its sensitivity to fuel recycling conditions

Tatematsu, Kenji; Sato, Osamu

JAERI-Research 2004-024, 35 Pages, 2005/01

JAERI-Research-2004-024.pdf:9.97MB

Many scenarios were defined for future development of nuclear power generation and fuel cycle systems in Japan. These scenarios were quantitatively analyzed from the viewpoint of plutonium recycling, natural uranium consumption, stock of spent fuel, etc. Following findings were obtained from the analysis. RMWRs will contribute to control the uranium consumption at certain finite levels if net conversion ratio (CR) is kept higher than 1.0. However, since RMWRs do not have an excellent breeding performance in comparison with FBRs, their effect is very sensitive to the conditions on fuel recycling processes. Judging from the results of analysis using a RMWR design with gross CR 1.06, it would be necessary for RMWRs to have net CR 1.04 in order to replace enriched uranium fuelled LWRs by around the year 2200, and thereby to keep ultimate natural uranium consumption at rather low levels. This can be achieved by controlling fuel duration time outside reactors to shorter than 4 years or 6 years, when total loss of plutonium during the processes of recycling is 1.0% or 0.2%, respectively.

JAEA Reports

Proceedings of the Symposium on Nitride Fuel Cycle Technology; July 28, 2004, JAERI, Tokai, Japan

Department of Materials Science

JAERI-Conf 2004-015, 143 Pages, 2004/12

JAERI-Conf-2004-015.pdf:17.09MB

This report is the Proceedings of the Symposium of Nitride Fuel Cycle Technology, which was held on July 28, 2004, at Tokai Research Establishment of Japan Atomic Energy Research Institute (JAERI). The purpose of this symposium is to exchange information and views on nitride fuel cycle technology among researchers from foreign and domestic organizations, and to discuss the recent and future research activities. The topics discussed in the symposium are Present State of the Technology Development in the World and Japan, Fabrication Technology, Property Measurement and Pyrochemical Process. The intensive discussion was made among 53 participants.

Journal Articles

Research on PARC process for future reprocessing

Asakura, Toshihide; Hotoku, Shinobu; Ban, Yasutoshi; Matsumura, Masakazu*; Kim, S.-Y.; Mineo, Hideaki; Morita, Yasuji

Proceedings of International Conference ATALANTE 2004 Advances for Future Nuclear Fuel Cycles, 5 Pages, 2004/06

In JAERI, PARC process based on PUREX technique has been studied to as the basis of future reprocessing. The key of concept is to obtain the products, U and Pu, within only a single extraction cycle by separating Np and Tc from U and Pu before U/Pu partition. Two flow-sheet tests on the process were performed with 44 GWd/t PWR spent-fuel solutions. It was demonstrated that remaining Np in raffinate from co-extraction could be decreased to 13 % compared to the dissolver solution with increased solvent flow rate and with increased nitric acid concentration of FP scrubbing solution. It was demonstrated that Np separation (selective reduction by n-butyraldehyde) efficiency could be improved from 36 % to 78 % by flow-sheet modification; increasing reductant concentration and scrubbing solution flow rate. The feasibility of the Tc separation technique by high acid scrubbing was demonstrated.

JAEA Reports

Development of testing techniques to evaluate thermal deformation behavior of fuel cladding tubes (Contract research)

Kaneko, Tetsuji; Tsukatani, Ichiro; Kiuchi, Kiyoshi

JAERI-Tech 2004-035, 18 Pages, 2004/03

JAERI-Tech-2004-035.pdf:0.81MB

Fuel elements used in the Reduced-Moderation Water Reactor (RMWR) have the stacking structure consisting of MOX pellets and UO$$_{2}$$ blankets in a fuel rod in order to attain the high breeding ratio and high burn-up simultaneously. It is a characteristic of the fuel elements that there is high thermal stress caused by inhomogeneous linear power density along the longitudinal direction of the fuel rod in comparison with the present LWR fuels. For this reason, it is important to estimate local deformation behavior of the fuel cladding tube with temperature difference caused by MOX pellet and UO$$_{2}$$ blanket. The testing machine was designed to investigate thermal-fatigue behavior under biaxial stress condition. The testing machine consists of the temperature distribution control unit, low cycle fatigue testing unit and internal pressure loading unit, it is also possible to conduct the simulation tests to investigate effects of pressure change with burn-up and longitudinal load change due to operation modes and restriction of fuel rods.

JAEA Reports

Progress of nuclear safety research, 2003

Editorial Committee on Nuclear Safety Research Results

JAERI-Review 2004-010, 155 Pages, 2004/03

JAERI-Review-2004-010.pdf:16.43MB

JAERl is conducting nuclear safety research primarily at the Nuclear Safety Research Center in close cooperation with the related departments in accordance with the Long Term Plan for Development and Utilization of Nuclear Energy and Annual Plan for Safety Research issued by the Japanese government. The fields of conducting safety research at JAERl are the engineering safety of nuclear power plants and nuclear fuel cycle facilities, and radioactive waste management as well as advanced technology for safety improvement or assessment. Also, JAERI has conducted international collaboration to share the information on common global issues of nuclear safety and to supplement own research. This report summarizes the nuclear safety research activities of JAERI from April 2001 through March 2003 and utilized facilities.

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