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Amaya, Masaki
High Temperature Corrosion of Materials, 15 Pages, 2024/00
Times Cited Count：0 Percentile：0.02(Metallurgy & Metallurgical Engineering)Narukawa, Takafumi; Kondo, Keietsu; Fujimura, Yuki; Kakiuchi, Kazuo; Udagawa, Yutaka; Nemoto, Yoshiyuki
Journal of Nuclear Materials, 587, p.154736_1 - 154736_8, 2023/12
Times Cited Count：1 Percentile：68.31(Materials Science, Multidisciplinary)Collaborative Laboratories for Advanced Decommissioning Science; The University of Tokyo*
JAEA-Review 2022-036, 115 Pages, 2023/01
JAEA-Review-2022-036.pdf:7.15MB
The Collaborative Laboratories for Advanced Decommissioning Science (CLADS), Japan Atomic Energy Agency (JAEA), had been conducting the Nuclear Energy Science & Technology and Human Resource Development Project (hereafter referred to "the Project") in FY2021. The Project aims to contribute to solving problems in the nuclear energy field represented by the decommissioning of the Fukushima Daiichi Nuclear Power Station (1F), Tokyo Electric Power Company Holdings, Inc. (TEPCO). For this purpose, intelligence was collected from all over the world, and basic research and human resource development were promoted by closely integrating/collaborating knowledge and experiences in various fields beyond the barrier of conventional organizations and research fields. The sponsor of the Project was moved from the Ministry of Education, Culture, Sports, Science and Technology to JAEA since the newly adopted proposals in FY2018. On this occasion, JAEA constructed a new research system where JAEA-academia collaboration is reinforced and medium-to-long term research/development and human resource development contributing to the decommissioning are stably and consecutively implemented. Among the adopted proposals in FY2020, this report summarizes the research results of the "Investigation of environment induced property change and cracking behavior in fuel debris" conducted in FY2021. The present study aims to investigate the environment induced property change and cracking behavior in fuel debris from the viewpoints of materials science. The research objective is cracking behavior in fuel debris which is presumed to be influenced by environment during long-term fuel debris processing period. The degradation models will be established to simulate the oxidation and hydrogenation processes possibly occurred at fuel debris. The evolution of phase constitution and the corresponding property change in the simulated fuel debris under various environmental conditions …
Collaborative Laboratories for Advanced Decommissioning Science; The University of Tokyo*
JAEA-Review 2021-058, 75 Pages, 2022/02
JAEA-Review-2021-058.pdf:4.82MB
The Collaborative Laboratories for Advanced Decommissioning Science (CLADS), Japan Atomic Energy Agency (JAEA), had been conducting the Nuclear Energy Science & Technology and Human Resource Development Project (hereafter referred to "the Project") in FY2020. The Project aims to contribute to solving problems in the nuclear energy field represented by the decommissioning of the Fukushima Daiichi Nuclear Power Station, Tokyo Electric Power Company Holdings, Inc. (TEPCO). For this purpose, intelligence was collected from all over the world, and basic research and human resource development were promoted by closely integrating/collaborating knowledge and experiences in various fields beyond the barrier of conventional organizations and research fields. The sponsor of the Project was moved from the Ministry of Education, Culture, Sports, Science and Technology to JAEA since the newly adopted proposals in FY2018. On this occasion, JAEA constructed a new research system where JAEA-academia collaboration is reinforced and medium-to-long term research/development and human resource development contributing to the decommissioning are stably and consecutively implemented. Among the adopted proposals in FY2020, this report summarizes the research results of the "Investigation of environment induced property change and cracking behavior in fuel debris" conducted in FY2020. The present study aims to investigate the environment induced property change and cracking behavior in fuel debris from the viewpoints of materials science. The research objective is cracking behavior in fuel debris which is presumed to be influenced by environment during long-term fuel debris processing period. The degradation models will be established to simulate the oxidation and hydrogenation processes possibly occurred at fuel debris.
Pham, V. H.; Kurata, Masaki; Steinbrueck, M.*
Thermo (Internet), 1(2), p.151 - 167, 2021/09
Nagase, Fumihisa; Narukawa, Takafumi; Amaya, Masaki
JAEA-Review 2020-076, 129 Pages, 2021/03
JAEA-Review-2020-076.pdf:3.9MB
Each light-water reactor (LWR) is equipped with the Emergency Core Cooling System (ECCS) to maintain the coolability of the reactor core and to suppress the release of radioactive fission products to the environment even in a loss-of-coolant accident (LOCA) caused by breaks in the reactor coolant pressure boundary. The acceptance criteria for ECCS have been established in order to evaluate the ECCS performance and confirm the sufficient safety margin in the evaluation. The limits defined in the criteria were determined in 1975 and reviewed based on state-of-the-art knowledge in 1981. Though the fuel burnup extension and necessary improvements of cladding materials and fuel design have been conducted, the criteria have not been reviewed since then. Meanwhile, much technical knowledge has been accumulated regarding the behavior of high-burnup fuel during LOCAs and the applicability of the criteria to the high-burnup fuel. This report provides a comprehensive review of the history and technical bases of the current criteria and summarizes state-of-the-art technical findings regarding the fuel behavior during LOCAs. The applicability of the current criteria to the high-burnup fuel is also discussed.
Narukawa, Takafumi; Amaya, Masaki
Journal of Nuclear Science and Technology, 57(1), p.68 - 78, 2020/01
Times Cited Count：2 Percentile：21.22(Nuclear Science & Technology)Narukawa, Takafumi; Amaya, Masaki
Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.912 - 921, 2019/09
Narukawa, Takafumi; Amaya, Masaki
Journal of Nuclear Science and Technology, 56(7), p.650 - 660, 2019/07
Times Cited Count：11 Percentile：76.81(Nuclear Science & Technology)Yumura, Takanori; Amaya, Masaki
Annals of Nuclear Energy, 120, p.798 - 804, 2018/10
Times Cited Count：6 Percentile：52.24(Nuclear Science & Technology)Negyesi, M.; Amaya, Masaki
Journal of Nuclear Science and Technology, 54(10), p.1143 - 1155, 2017/10
Times Cited Count：6 Percentile：50.9(Nuclear Science & Technology)Amaya, Masaki; Udagawa, Yutaka; Narukawa, Takafumi; Mihara, Takeshi; Taniguchi, Yoshinori
Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 10 Pages, 2017/09
inert matrix fuelMiwa, Shuhei; Osaka, Masahiko
Journal of Nuclear Materials, 487, p.1 - 4, 2017/04
Times Cited Count：4 Percentile：36.71(Materials Science, Multidisciplinary)Oxidation and reduction behaviors of prototypic MgO-based inert matrix fuels (IMFs) containing PuO were experimentally investigated by means of thermogravimetry. The oxidation and reduction kinetics of the MgO-PuO specimen were determined. The oxidation and reduction rates of the MgO-PuO were found to be low compared with those of PuO. It is note that the changes in O/Pu ratios of MgO-PuO from stoichiometry were smaller than those of PuO at high oxygen partial pressure. From these results, it can be said that MgO matrix lower the oxygen supply and release of PuO, which is preferable as the minor actinides incineration devices, since the high oxygen potentials of minor actinide oxides can cause certain problems in terms of thermochemical aspects such as enlarged cladding inner-surface corrosion.
Ueta, Shohei; Sumita, Junya; Shibata, Taiju; Aihara, Jun; Fujita, Ichiro*; Ohashi, Jun*; Nagaishi, Yoshihide*; Muto, Takenori*; Sawa, Kazuhiro; Sakaba, Nariaki
Nuclear Engineering and Design, 271, p.309 - 313, 2014/05
Times Cited Count：9 Percentile：57.01(Nuclear Science & Technology)A new concept of the high temperature gas-cooled reactor (HTGR) is proposed as a challenge to assure no event sequences to the harmful release of radioactive materials even when the design extension conditions (DECs) occur by deterministic approach based on the inherent safety features of the HTGR. The air/water ingress accident, one of the DECs for the HTGR, is prevented by additional measures (e.g. facility for suppression to air ingress). With regard to the core design, it is important to prevent recriticality accidents by keeping the geometry of the fuel rod which consists of the graphite sleeve, fuel compact and SiC-TRISO (TRIstructural-ISOtropic) coated fuel particle, and by improving the oxidation resistance of the graphite when air/water ingress accidents occur. Therefore, it is planned to develop the oxidation-resistant graphite, which is coated with gradient SiC layer. It is also planned that the experimental identification of the condition to form the stable oxide layer (SiO
) for SiC layer on the oxidation-resistant graphite and on the SiC-TRISO fuel. This paper describes the R&D plan for un-irradiation and irradiation test under simulating air/water ingress accident condition to develop oxidation-resistant graphite and to investigate the oxidation behavior of SiC coated fuel particle.
Nagase, Fumihisa; Fuketa, Toyoshi
Proceedings of 2004 International Meeting on LWR Fuel Performance, p.500 - 506, 2004/09
A systematic research program is being conducted at the Japan Atomic Energy Research Institute (JAERI), which aims at a wide range database for evaluating the influence of further burnup extension on fuel behavior under LOCA conditions. As a part of the program, integral thermal shock tests simulating the whole LOCA sequence have been conducted with Zircaloy-4 fuel claddings, irradiated to 39 and 44GWd/t at a PWR. One cladding, oxidized to about 30% ECR, fractured during the quench. The fracture condition agrees with the fracture criteria for non-irradiated claddings that have similar hydrogen concentrations (about 25% ECR). Two claddings, oxidized to about 16 and 18% ECR, survived the quench, indicating that fracture/non-fracture boundary is not reduced so significantly by irradiation for the examined burnup range. The present paper describes information obtained from the tests including oxidation kinetics and rupture behavior.
Nagase, Fumihisa; Fuketa, Toyoshi
Journal of Nuclear Science and Technology, 41(7), p.723 - 730, 2004/07
Times Cited Count：45 Percentile：92.54(Nuclear Science & Technology)Experiments simulating loss-of-coolant accident (LOCA) conditions were performed to evaluate effect of pre-hydriding on thermal-shock resistance of oxidized Zircaloy-4 cladding. Artificially hydrided (400 to 600 ppm) and non-hydrided claddings were subjected to the tests. Since cladding fracture on quenching primarily depends on the oxidation amount, fracture threshold was evaluated in terms of "Equivalent Cladding Reacted (ECR)". Under axially non-restrained condition, the fracture threshold is 56% ECR and the influence of pre-hydriding is not seen. The fracture threshold is decreased by restraining the test rods on quenching, and it is more remarkable in pre-hydrided claddings than in non-hydrided claddings. Consequently, the fracture threshold was 20% ECR and 10% ECR for non-hydrided and pre-hydrided claddings, respectively, under the fully restrained condition. These results indicate possible decrease of fracture threshold of high burnup fuel claddings under LOCA conditions.
Nagase, Fumihisa; Fuketa, Toyoshi
NUREG/CP-0185, p.321 - 331, 2004/00
With a view to obtaining basic data to evaluate high burnup fuel behavior under loss of coolant accident (LOCA) conditions, a research program is being conducted at the Japan Atomic Energy Research Institute (JAERI). The program consists of integral thermal shock tests and other separate tests for oxidation rate and mechanical property of fuel claddings. Prior to the tests on irradiated claddings, the tests have been conducted on non-irradiated claddings to examine separate effects of corrosion and hydrogen absorption during reactor operation. Hydrogen effects have been especially examined because hydrogen absorption has the great impact on cladding embrittlement. The tests on irradiated claddings have recently been started and preliminary results have been obtained. The present paper summarizes recent results from those studies.
Nagase, Fumihisa; Uetsuka, Hiroshi
NUREG/CP-0176, p.335 - 342, 2002/05
no abstracts in English
Boyack, B. E.*; Motta, A. T.*; Peddicord, K. L.*; Alexander, C. A.*; Andersen, J. G. M.*; Blaisdell, J. A.*; Dunn, B. M.*; Ebeling-Koning, D.*; Fuketa, Toyoshi; Hache, G.*; et al.
NUREG/CR-6744, 455 Pages, 2001/12
no abstracts in English
Boyack, B. E.*; Motta, A. T.*; Peddicord, K. L.*; Andersen, J. G. M.*; Alexander, C. A.*; Dunn, B. M.*; Fuketa, Toyoshi; Hochreiter, L. E.*; Montgomery, R. O.*; Moody, F. J.*; et al.
NUREG/CR-6743, 266 Pages, 2001/09
no abstracts in English
