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Pu
O
powder using master sintering curve theoryNakamichi, Shinya; Sunaoshi, Takeo*; Hirooka, Shun; Vauchy, R.; Murakami, Tatsutoshi
Journal of Nuclear Materials, 595, p.155072_1 - 155072_11, 2024/07
Oizumi, Akito; Akie, Hiroshi
JAEA-Technology 2023-017, 93 Pages, 2023/12
JAEA-Technology-2023-017.pdf:8.45MB
After the decision of decommissioning JMTR (Japan Materials Testing Reactor), Japan Atomic Energy Agency investigated the possibility to construct a new irradiation test reactor to succeed JMTR (post-JMTR), and the final report of the investigated result was submitted to the Ministry of Education, Culture, Sports, Science and Technology on March 30th 2021. This investigation was carried out in 4 steps of (1) selection of reactor type, (2) reactor core plans studies, (3) neutronic studies, (4) thermal studies, and was finally (5) considered and evaluated. This JAEA-Technology report summarizes the process and the results of (3) neutronic studies. Neutron fluxes were calculated at irradiation sample positions in the investigated cores, the standard core and the compact core, and the calculated fluxes satisfied the required irradiation capability. It was also evaluated the two investigated cores' continuous reactor operation time in days in one refueling cycle, and the results guaranteed an operation days equality with that of existing JMTR. In addition, neutronic characteristics of the cores were estimated, such as power distribution in the core, control rod reactivity worth, reactivity coefficients, distribution of fuel burnup rate of each fuel element, and kinetics parameters. The evaluated neutronic characteristics were used in the post-JMTR final investigation report to confirm the neutronic feasibility by comparing with the neutronic limiting values of existing JMTR, and to estimate the cooling capability to make the core thermally feasible.
Yamane, Yuichi
Journal of Nuclear Science and Technology, 59(11), p.1331 - 1344, 2022/11
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)The reactivity was estimated from a time profile of neutron count rate or a simulated data in a quasi-steady state after sudden change of reactivity or external neutron source strength. The estimation was based on the equation of power in subcritical quasi-steady state. The purpose of the study is to develop the method of timely reactivity estimation from complicated time profile of neutron count rate. The developed method was applied to the data simulating neutron count rate created by using one-point kinetics code, AGNES, and Poisson-distributed random noise and to the transient subcritical experiment data measured by using TRACY. The result shows that the difference of the estimated and reference value was within about 5% or less for (
-1) for simulated data and within about 7% or less for 
-1.4 and -3.1 for the experimental data. It was also shown that the possibility of the reactivity estimation several ten seconds after the status change.
Yoshida, Naoki; Ono, Takuya; Yoshida, Ryoichiro; Amano, Yuki; Abe, Hitoshi
JAEA-Research 2021-011, 12 Pages, 2022/01
JAEA-Research-2021-011.pdf:1.49MB
In boiling and drying accidents involving high-level liquid waste in fuel reprocessing plants, emphasis is placed on the behavior of ruthenium (Ru). Ru would form volatile species, such as ruthenium tetroxide (RuO
), and could be released to the environment with coexisting gases, including nitric acid, water, or nitrogen oxides. In this study, to contribute toward safety evaluations of these types of accidents, the migration behavior of gaseous Ru into the liquid phase has been experimentally measured by simulating the condensate during an accident. The gas absorption of RuO was enhanced by increasing the nitrous acid (HNO) concentration in the liquid phase, indicating the occurrence of chemical absorption. In control experiments without HNO, the lower the temperature, the greater was the Ru recovery ratio in the liquid phase. Conversely, in experiments with HNO, the higher the temperature, the higher the recovery ratio, suggesting that the reaction involved in chemical absorption was activated at higher temperatures.
Yoshida, Naoki; Amano, Yuki; Ono, Takuya; Yoshida, Ryoichiro; Abe, Hitoshi
JAEA-Research 2020-014, 33 Pages, 2020/12
JAEA-Research-2020-014.pdf:3.66MB
Considering the boiling and drying accident of high-level liquid waste in fuel reprocessing plant, Ruthenium (Ru) is an important element. It is because Ru would form volatile compounds such as ruthenium tetroxide (RuO) and could be released into the environment with other coexisting gasses such as nitric oxides (NOx) such as nitric oxide (NO) and nitrogen dioxide (NO). To contribute to the safety evaluation of this accident, we experimentally evaluated the effect of NOx on the decomposition and chemical change behavior of the gaseous RuO (RuO(g)). As a result, the RuO(g) decomposed over time under the atmospheric gasses with NO or NO, however, the decomposition rate was slower than the results of experiments without NOx. These results showed that the NOx stabilized RuO(g).
Yamane, Yuichi
Journal of Nuclear Science and Technology, 57(8), p.926 - 931, 2020/08
Times Cited Count：1 Percentile：12.16(Nuclear Science & Technology)An equation of power in subcritical quasi-steady state has been derived based on one-point kinetics equations for the purpose of utilizing it for the development of timely reactivity estimation from complicated time profile of neutron count rate. It linearly relates power, 
, to a new variable 
, which is a function of time differential of the power. It has been confirmed by using one-point kinetics code, AGNES, that the calculated points (
) are perfectly in a line described by the new equation and that points () calculated from transient subcritical experiments by using TRACY made a line with a slope indicated by the new equation.
Aoki, Takeshi; Chirayath, S. S.*; Sagara, Hiroshi*
Annals of Nuclear Energy, 141, p.107325_1 - 107325_7, 2020/06
Times Cited Count：2 Percentile：24.28(Nuclear Science & Technology)The proliferation resistance (PR) of an inert matrix fuel (IMF) in the transuranic nuclear fuel cycle (NFC) of a high temperature gas cooled reactor is evaluated relative to the uranium and plutonium mixed-oxide (MOX) NFC of a light water reactor using PRAETOR code and sixty-eight input attributes. The objective is to determine the impacts of chemical stability of IMF and fuel irradiation on the PR. Specific material properties of the IMF, such as lower plutonium content, carbide ceramics coating, and absence of 
U, contribute to enhance its relative PR compared to MOX fuel. The overall PR value of the fresh IMF (an unirradiated direct use material with a one-month diversion detection timeliness goal) is nearly equal to that of the spent MOX fuel (an irradiated direct use nuclear material with a three-month diversion detection timeliness goal). Final results suggest a reduced safeguards inspection frequency to manage the IMF.
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.
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.
Takeuchi, Masayuki; Sano, Yuichi; Watanabe, So; Nakahara, Masaumi; Aihara, Haruka; Kofuji, Hirohide; Koizumi, Tsutomu; Mizuno, Tomoyasu
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 6 Pages, 2017/04
Aihara, Haruka; Arai, Yoichi; Shibata, Atsuhiro; Nomura, Kazunori; Takeuchi, Masayuki
Procedia Chemistry, 21, p.279 - 284, 2016/12
Times Cited Count：5 Percentile：94.32(Chemistry, Inorganic & Nuclear)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.
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.
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 
-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 
Am in pyrochemical process was investigated by electrochemical methods.
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
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.
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.
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.
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.
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.
