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Ikusawa, Yoshihisa; Nagayama, Masahiro*
JAEA-Data/Code 2023-006, 24 Pages, 2023/07
JAEA-Data-Code-2023-006.pdf:1.42MB
Core fuels with stainless steel cladding and high plutonium content mixed oxide (MOX) fuel in a water-cooled environment, such as supercritical water-cooled reactors (SCWR) and reduced-moderation water reactors (RMWR), have been studied. In order to contribute to the research and development of such a core fuel concept, the fuel performance code "FEMAXI-8" was verified based on the results of post irradiation examinations of MOX fuel irradiated in the experimental fast reactor "JOYO". FEMAXI-8 is the latest version of the behavior analysis code developed by JAEA to analyze the behavior of light water reactor fuels under normal operation and transient conditions. This latest code has been improved and developed to allow the selection of stainless steel cladding property models to analyze improved fuels such as accident tolerant fuels. The purpose of this report is to confirm the prediction accuracy of FEMAXI-8 for the irradiation behavior of the new type of core fuel that is currently being developed. As a result of the verification, it was confirmed that FEMAXI-8 has sufficient analysis accuracy for the irradiation behavior of sodium-cooled fast reactor MOX fuel with stainless steel cladding, which exceeds the plutonium content and irradiation conditions of light water reactors. In the future, the analysis accuracy of FEMAXI-8 could be improved by adopting the O/M ratio dependence of MOX fuel thermal conductivity and the irradiation behavior evaluation model at high temperature.
Yamashita, Shinichiro
Nihon Genshiryoku Gakkai-Shi ATOMO
, 65(4), p.233 - 237, 2023/04
In the wake of the accident at the Fukushima Daiichi Nuclear Power Plant (NPP) of TEPCO due to the Great East Japan Earthquake in 2011, interest in the early implementation of accident tolerant fuel (ATF) not only for many existing NPPs but also for future NPPs, which is expected to dramatically improve the safety of light water reactors, has increased globally, and research and development is currently underway in many countries around the world. In this article, an overview of domestic ATF technology development that has been carried out with the support of the Ministry of Economy, Trade and Industry since 2015, will be introduced.
Nagase, Fumihisa
Annals of Nuclear Energy, 171, p.109052_1 - 109052_8, 2022/06
Times Cited Count：2 Percentile：53.91(Nuclear Science & Technology)The fracture threshold of the fuel decreases if the oxidized Zr alloy cladding is strongly constrained by the spacer grid during quenching in a loss-of-coolant accident. Therefore, the estimation of realistic levels of the axial constraint has been a subject of significant interest on fuel safety. In this study, a test assembly consisting of a PWR-type simulated fuel segment and a 3
3 grid piece was heated in steam, cooled, and quenched, and the axial constraint force on the fuel segment was measured. The constraint force of the Zircaloy grid gradually decreased with temperature. Once the Zircaloy grid was heated to 
1060 K, the reduced constraint force had difficulty recovering, and thus the maximum constraint force during cooling and quenching was 
10 N. The constraint force was clearly reduced at 1070 K during the tests with the Inconel grid. However, the reduced constraint force partially recovered during cooling. As a result, the maximum constraint force during cooling and quenching was 20 to 50 N for the Inconel grid. In conclusion, oxidation, ballooning, rupture, or eutectic formation would not generally cause an extremely strong constraint, as predicted by previous studies, at the grid position.
Ozawa, Masaaki*; Amaya, Masaki
Nihon Genshiryoku Gakkai Wabun Rombunshi, 19(4), p.185 - 200, 2020/12
no abstracts in English
Fukaya, Yuji
Annals of Nuclear Energy, 144, p.107503_1 - 107503_7, 2020/09
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Disposal scenario of High Level Waste (HLW) of Light Water Reactor (LWR) have been optimized to reduce waste volume and repository footprint in geological disposal. The optimization was performed with existing and demonstrated technology in the situation where the reprocessing will be delayed. In general, the scenario with Partitioning and Transmutation (P&T) is optimized to minimize waste package number generated in the situation where the spent fuel will be reprocessed immediately with the minimum cooling time. With considering the delay of reprocessing, it is found that the more simplified and effective optimization with the high-waste-loading glass and cold crucible induction melter technologies and without partitioning. The optimized case can achieve significant reduction of number of waste package generation and the repository footprint to half of those of non-P&T case with 100 years cooling.
Kasahara, Shigeki; Chimi, Yasuhiro; Hata, Kuniki; Hanawa, Satoshi
Zairyo To Kankyo, 68(9), p.240 - 247, 2019/09
In order to study environment assisted cracking mechanism of stainless steel under BWR primary coolant condition, effects of applied load on oxidation in the vicinity of crack tips of CT specimens were evaluated. Loaded CT specimens were immersed in an aqueous condition at 290
C as a simulated BWR coolant condition, and microstructural observation on oxide near the tips of pre-cracks was carried out. Oxide inner layers, which consisted of fine grain magnetite containing Fe and Cr were formed, and oxide outer layers consisting of large grains of Fe
O
were observed to cover the inner layers. FEM analysis of stress and strain in the loaded CT specimen suggests that both of dislocations due to localized plastic deformation and elastic strain could play important roles to accelerate inner oxide formation in the vicinity of the crack tip of the specimens.
Hino, Ryutaro; Takegami, Hiroaki; Yamazaki, Yukie; Ogawa, Toru
JAEA-Review 2016-038, 294 Pages, 2017/03
JAEA-Review-2016-038.pdf:11.08MB
In the aftermath of the Fukushima nuclear accident, safety measures against hydrogen in severe accident have been recognized as a serious technical problem in Japan. Therefore, efforts have begun to form a common knowledge base between nuclear engineers and experts on combustion and explosion, and to secure and improve future nuclear energy safety. As one of such activities, we have prepared the "Handbook of Advanced Nuclear Hydrogen Safety" under the Advanced Nuclear Hydrogen Safety Research Program funded by the Agency for Natural Resources and Energy of the Ministry of Economy, Trade and Industry. The concepts of the handbook are as follows: to show advanced nuclear hydrogen safety technologies that nuclear engineers should understand, to show hydrogen safety points to make combustion-explosion experts cooperate with nuclear engineers, to expand information on water radiolysis considering the situation from just after the Fukushima accidents and to the waste management necessary for decommissioning after the accident, etc.
Yamamoto, Masahiro
Zairyo To Kankyo, 66(1), p.3 - 12, 2017/01
The laboratory simulation tests which could be reproduced the corrosion reactions propagating in the actual environments were utilized to analyze the mechanism of corrosion phenomena. In this report, some results are introduced in the cases of maritime structures and nuclear facilities. Experimental apparatus was originally designed to obtain the data in high radioactive condition simulating actual plants. One is a result showing the effect of Np ion to the corrosion of stainless steel in nuclear fuel reprocessing plant. Corrosion mechanism was revealed that Np
ion is reduced to Np
ion by a corrosion reaction of stainless steel and then re-oxidized to Np ion in the bulk solution. And repetition of this cycle accelerated corrosion of stainless steel by a little amounts of Np addition in nitric acid solution. Another result is introduced that an effect of H
O created by radiolysis of cooling water at high radioactive environment in light water reactor.
Yamamoto, Keiichi; Takeuchi, Tomoaki; Hayashi, Takayasu*; Kosuge, Fumiaki*; Tsuchiya, Kunihiko
JAEA-Testing 2016-002, 25 Pages, 2016/11
JAEA-Testing-2016-002.pdf:4.97MB
Development of the reactor measurement system has been carried out to obtain the real-time in-core nuclear and thermal information, where the quantitative measurement of brightness of Cherenkov light was investigated. The system would be applied as a monitoring system in severe accidents and for the advanced operation management technology in existing LWRs. This report summarized the modification of Cherenkov light estimation system described JAEA-Testing 2015-001 and the result of the burn-up evaluation by Cherenkov light image emitted from spent fuel elements of LWRs with the modified system.
Yamashita, Shinichiro; Nagase, Fumihisa; Kurata, Masaki; Kaji, Yoshiyuki
Proceedings of Annual Topical Meeting on LWR Fuels with Enhanced Safety and Performance (TopFuel 2016) (USB Flash Drive), p.21 - 30, 2016/09
Fuel rod, channel box, and control rod designed with new materials and concepts have been developed in Japan for increasing accident tolerance of LWRs. In order to efficiently and properly implement the accident tolerant fuels (ATFs) and the other components, it is necessary not only to accumulate fundamental and practical data but also to consider technology readiness, recognize knowledge gaps, and establish strategy for design and fabrication. The Japan Atomic Energy Agency (JAEA) has established the above "technical basis" and drafted a research plan towards implementation of the ATFs and components as a program sponsored and organized by the Ministry of Economy, Trade and Industry (METI). It is useful to take advantage of the experiences in commercial uses of zirconium-base alloys in LWRs and, therefore, JAEA has conducted this METI project in cooperation with power plant providers, fuel venders, research institutes and universities who have been involved in the development of the ATF materials. The present paper describes the main results of the project conducted to establish the technical basis of the ATFs and components.
Tokai Reprocessing Technology Development Center
JAEA-Evaluation 2015-012, 83 Pages, 2015/12
JAEA-Evaluation-2015-012.pdf:6.67MB
Japan Atomic Energy Agency (hereafter referred as "JAEA") consulted the "Evaluation Committee of Research and Development Activities for Fast Reactor Cycle" to assess the issue on "Research and Development on Reprocessing of Nuclear Fuel Materials" conducted by JAEA during the period from FY2010 to FY2014. In response to the JAEA's request, the committee assessed the R&D programs and the activities of JAEA related to the issue and concluded the mission was accomplished. This evaluation was performed based on the "General guideline for the evaluation of government R&D activities", the "Guideline for evaluation of R&D in Ministry of Education, Culture, Sports, Science and Technology (MEXT)" and the "Operational rule for evaluation of R&D activities" by JAEA.
Fuketa, Toyoshi; Sugiyama, Tomoyuki; Sasajima, Hideo; Nagase, Fumihisa
Proceedings of 2005 Water Reactor Fuel Performance Meeting (CD-ROM), p.633 - 645, 2005/10
LWR fuel behaviors during a reactivity initiated accident (RIA) are being studied in the NSRR program. Results from recent NSRR experiments, no failures in Tests OI-10 and -12 and the higher failure enthalpy in Test OI-11, reflect the better performance of the new cladding materials in terms of corrosion during PWR operations. Accordingly, these rods with improved corrosion resistance have larger safety margin than conventional Zircaloy-4 rods. In addition, the smaller inventory of inter-granular gas in the large grain pellet could reduce the fission gas release in RIA as observed in the OI-10. Test VA-1 was conducted with an MDA sheathed 78 MWd/kgU PWR fuel rod. Despite of the higher burnup and thicker oxide layer of 
81
m, the enthalpy at failure remained in a same level as those for rods with of 40m-oxide at 50 - 60 MWd/kgU. This result suggests high burnup structure (rim structure) in pellet periphery does not have strong effect on the failure enthalpy reduction because the PCMI load is produced primarily by solid thermal expansion of the pellet.
Ueno, Fumiyoshi; Nagae, Yuji*; Nemoto, Yoshiyuki; Miwa, Yukio; Takaya, Shigeru*; Hoshiya, Taiji*; Tsukada, Takashi; Aoto, Kazumi*; Ishii, Toshimitsu; Omi, Masao; et al.
JAERI-Research 2005-023, 132 Pages, 2005/09
JAERI-Research-2005-023.pdf:33.03MB
JAERI and JNC have started a JAERI-JNC joint research program in fiscal year 2003, which has been aimed for efficient progress and synergistic effect on the research activities in both Institutes. This study has been chosen one of the joint research themes because it has been our common objective in the field of structural materials of FBR and LWR components. The purpose of the study is to clarify damage mechanism of structural materials used under irradiation, and then to develop the methods for damage evaluation and detection in earlier stage of progressing process of damage. In fiscal year 2004 and 2005, micro-corrosion measurement, electrochemical corrosion test and leakage magnetic flux density measurement apparatuses were developed and equipped in two hot facilities and irradiated and unirradiated crept specimens, irradiated high purity model austenitic stainless alloys were also prepared and applied to this study. These apparatuses and specimens were used for damage evaluation, and these feasibilities for nuclear power plant materials were studied.
Akie, Hiroshi; Nabeshima, Kunihiko; Uchikawa, Sadao
JAERI-Conf 2005-009, 153 Pages, 2005/08
JAERI-Conf-2005-009.pdf:14.7MB
As a research on the future innovative water reactor, the development of Reduced-Moderation Water Reactors (RMWRs) has been performed in JAERI. The workshop on RMWRs is aiming at information exchange between JAERI and other organizations, and has been held every year since 1998. The program of the 7th workshop was composed of 5 lectures and an overall discussion time. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture as well as of the discussion time. In addition in Appendix, there are included presentation handouts of each lecture.
Yoshida, Hiroyuki; Ose, Yasuo*; Kureta, Masatoshi*; Nagayoshi, Takuji*; Takase, Kazuyuki; Akimoto, Hajime
Nihon Genshiryoku Gakkai Wabun Rombunshi, 4(2), p.106 - 114, 2005/06
no abstracts in English
Yoshida, Hiroyuki; Nagayoshi, Takuji*; Ose, Yasuo*; Takase, Kazuyuki; Akimoto, Hajime
Nihon Genshiryoku Gakkai Wabun Rombunshi, 4(1), p.25 - 31, 2005/03
no abstracts in English
Sugiyama, Tomoyuki; Fuketa, Toyoshi
Journal of Nuclear Science and Technology, 41(11), p.1083 - 1090, 2004/11
Times Cited Count：10 Percentile：55.72(Nuclear Science & Technology)The effect of cladding surface pre-oxidation on the rod coolability under reactivity initiated accidents was investigated. NSRR tests on irradiated fuel rods have shown higher rod coolability than that of fresh rods, which arose from suppressed DNB and early quench at the surface. To identify the dominant factor, possible factors such as pellet cracking and so on, were assessed. The most probable factor, the cladding pre-oxidation, was examined by pulse irradiation tests on fresh rods with three cladding surface conditions, no oxide layer, 1m and 10m-thick oxide layers. Temperature measurements showed increased thresholds for DNB and quench at the pre-oxidized surface, leading to a reduced film boiling duration. The shifts of the critical and minimum heat flux points could be caused by the surface wettability increase. In the present tests, the wettability change was probably dominated by the chemical potential change at the surface due to pre-oxidation. The test results indicate the effects do not depend on the oxide layer thickness, but on the presence of the oxide layer.
Hoshiya, Taiji*; Ueno, Fumiyoshi; Takaya, Shigeru*; Nagae, Yuji*; Nemoto, Yoshiyuki; Miwa, Yukio; Aoto, Kazumi*; Tsukada, Takashi; Abe, Yasuhiro*; Nakamura, Yasuo*; et al.
JAERI-Research 2004-016, 53 Pages, 2004/10
JAERI-Research-2004-016.pdf:22.07MB
Japan Nuclear Cycle Development Institute (JNC) and Japan Atomic Energy Research Institute (JAERI) have started a JNC-JAERI united research program cooperatively in 2003, which has been aimed for efficient progress and synergistic effect on the research activities of both Institutes in order to lead the facing task of unification between JNC and JAERI. This study has been chosen one of the united research themes, and the purpose of it is to clarify damage mechanism of structural materials under irradiation, and then to develop the methods for damage evaluation and detection in earlier stage of progressing process of damage. In fiscal year 2003, magnetic flux density distribution (JNC) and micro-corrosion (JAERI) measurement apparatus were newly developed and equipped in Hot Facilities in two Institutes, respectively. These apparatus were designed and produced in consideration of radiation resistance and remote-controlled operation to equip in hot cells. We will start the study on neutron irradiation damage by employing the two apparatus as the next step.
Onuki, Akira
Nihon Kikai Gakkai Doryoku Enerugi Shisutemu Bumon Nyusu Reta, (29), p.3 - 5, 2004/10
no abstracts in English
Kureta, Masatoshi
Kashika Joho Gakkai-Shi, 24(Suppl.1), p.265 - 268, 2004/07
Visualization of 3D and instantaneous void fraction distribution of boiling flow in a tight-lattice 14-rod bundle is conducted by using neutron tomography and high-frame- rate neutron radiography void fraction measurement techniques. The purpose of the experiment is to understand vapor bubbles/water behavior ranging from the onset of boiling to the high void fraction region based on ("3D" + "2D+Time") void fraction data, and to obtain the fine-mesh database for verification of advanced analysis codes. Following phenomena are made clear from the present experiment: Vapor accumulates in the channel center; High void fraction spots appear between adjacent heater rods, that is, in narrow space at the inlet; Void fraction in the triangular space among three rods becomes high by void drift phenomenon, and "vapor chimney" is formed; Flow is intermittent, and vapor bubble clusters are formed periodically; Onset points of net vapor generation are scattered not only in the center but in the peripheral.
