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Li, X.; Yamaji, Akifumi*; Sato, Ikken*; Yamashita, Takuya
Annals of Nuclear Energy, 214, p.111217_1 - 111217_13, 2025/05
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Yamada, Takeshi*; Li, X.; Yamashita, Takuya; Yamaji, Akifumi*
Proceedings of 31st International Conference on Nuclear Engineering (ICONE31) (Internet), 10 Pages, 2024/11
In this study, a new crust model is being developed to analyze MCCI, which involves continuous concrete ablation with presence of the crust layer between the corium and the concrete walls, which may gradually move with the slow concrete wall ablation process over long time. The new crust model must enable accumulation of physical displacement of the crust particle over long time (i.e., enable physical creeping) while preventing accumulation of numerical displacement of the crust particles over long time (i.e., preventing numerical creeping), Hence, in the new crust model, the PS has been effectively disabled for the crust particles. Qualitative validity of such numerical modeling was confirmed through some trial analyses of VULCANO-VBS test using a set of tentative calculation conditions and parameters, which should be carefully revised for future quantitative discussions including validation of the analysis results with experimental results.
Li, X.; Yamaji, Akifumi*; Sato, Ikken*; Yamashita, Takuya; Nagae, Yuji
Proceedings of 11th European Review Meeting on Severe Accident Research Conference (ERMSAR 2024) (Internet), 12 Pages, 2024/05
Yamashita, Takuya; Shimomura, Kenta; Nagae, Yuji; Yamaji, Akifumi*; Mizokami, Shinya; Mitsugi, Takeshi; Koyama, Shinichi
Hairo, Osensui, Shorisui Taisaku Jigyo Jimukyoku Homu Peji (Internet), 53 Pages, 2023/10
JAEA performed the subsidy program for the "Project of Decommissioning, Contaminated Water and Treated Water Management (Development of Analysis and Estimation Technologies for Characterization of Fuel Debris (Development of Estimation Technologies of RPV Damaged Condition, etc.) in 2022JFY. This presentation summarized briefly the results of the project, which will be available shortly on the website of Management Office for the Project of Decommissioning, Contaminated Water and Treated Water Management.
Bando, Yamato*; Yamaji, Akifumi*; Yamashita, Takuya
Proceedings of International Conference on Environmental Remediation and Radioactive Waste Management (ICEM2023) (Internet), 9 Pages, 2023/10
Sato, Ikken; Yamaji, Akifumi*; Li, X.*; Madokoro, Hiroshi
Mechanical Engineering Journal (Internet), 9(2), p.21-00436_1 - 21-00436_17, 2022/04
Yoshikawa, Shinji; Yamaji, Akifumi*
JAEA-Research 2021-006, 52 Pages, 2021/09
JAEA-Research-2021-006.pdf:3.89MB
In Fukushima Daiichi Nuclear Power Station (referred to as "FDNPS" hereafter) unit2 and unit3, failure of the reactor pressure vessel (RPV) and relocation of some core materials (CRD piping elements and upper tie plate, etc.) to the pedestal region have been confirmed. In boiling water reactors (BWRs), complicated core support structures and control rod drive mechanisms are installed in the RPV lower head and its upper and lower regions, so that the relocation behavior of core materials to pedestal region is expected to be also complicated. The Moving Particle Semi-implicit (MPS) method is expected to be effective in overviewing the relocation behavior of core materials in complicated RPV lower structure of BWRs, because of its Lagrangian nature in tracking complex interfaces. In this study, for the purpose of RPV ablation analysis of FDNPS unit2 and unit3, rigid body model, parallelization method and improved calculation time step control method were developed in FY 2019 and improvement of pressure boundary condition treatment, stabilization of rigid body model, and calculation cost reduction of debris bed melting simulation were achieved in FY2020. These improvements enabled sensitivity analyses of melting, relocation and re-distribution behavior of deposited solid debris in RPV lower head on various cases, within practical calculation cost. As a result of the analyses of FDNPS unit2 and unit3, it was revealed that aspect (particles/ingots) and distribution (degree of stratification) of solidified debris in lower plenum have a great impact on the elapsed time of the following debris reheat and partial melting and on molten pool formation process, further influencing RPV lower head failure behavior and fuel debris discharging behavior.
Wang, Z.; Duan, G.*; Koshizuka, Seiichi*; Yamaji, Akifumi*
Nuclear Power Plant Design and Analysis Codes, p.439 - 461, 2021/00
Yamaji, Akifumi*; Susuki, Naomichi*; Kaji, Yoshiyuki
IAEA-TECDOC-1921, p.199 - 209, 2020/07
The thermo-physical models and irradiation behavior of FeCrAl as defined by the benchmark organizer have been implemented to FEMAXI-7. Analyses were carried out firstly for the specified normal operation condition. Then, some sensitivity analyses were carried out with different assumptions and model parameters. Under the normal operating condition, the predicted FeCrAl cladded fuel performance was similar to that of Zry cladded fuel with notable, but not major difference regarding late gap closure. Under the simulated LOCA conditions, the burst pressure could be evaluated. The predicted cladding creep strain at burst was mainly attributed to creep strain with negligible plastic strain. Overall, FEMAXI-7 analyses have demonstrated excellent robustness and flexibility in modeling FeCrAl-UO
system under normal and LOCA conditions.
Li, X.; Sato, Ikken; Yamaji, Akifumi*
Annals of Nuclear Energy, 133, p.21 - 34, 2019/11
Times Cited Count：6 Percentile：47.17(Nuclear Science & Technology)Pastore, G.*; Gamble, K. A.*; Cherubini, M.*; Giovedi, C.*; Marino, A.*; Yamaji, Akifumi*; Kaji, Yoshiyuki; Van Uffelen, P.*; Veshchunov, M.*
Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.1038 - 1047, 2019/09
Oxidation-resistant iron-chromium-aluminum (FeCrAl) steels have been proposed for application as cladding materials in light water reactor fuel rods with improved accident tolerance. Within the Coordinated Research Project ACTOF of the International Atomic Energy Agency (IAEA), a fuel performance modeling benchmark for FeCrAl cladding behavior was conducted. During this effort, calculations were performed with various fuel performance codes for a set of fuel rod problems with FeCrAl steel as cladding material, and results were compared to each other.
Yamashita, Shinichiro; Ioka, Ikuo; Nemoto, Yoshiyuki; Kawanishi, Tomohiro; Kurata, Masaki; Kaji, Yoshiyuki; Fukahori, Tokio; Nozawa, Takashi*; Sato, Daiki*; Murakami, Nozomu*; et al.
Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.206 - 216, 2019/09
After the nuclear accident at Fukushima Daiichi Power Plant, research and development (R&D) program for establishing technical basis of accident-tolerant fuel (ATF) started from 2015 in Japan. Since then, both experimental and analytical studies necessary for designing a new light water reactor (LWR) core with ATF candidate materials are being conducted within the Japanese ATF R&D Consortium for implementing ATF to the existing LWRs, accompanying with various technological developments required. Until now, we have accumulated experimental data of the candidate materials by out-of-pile tests, developed fuel evaluation codes to apply to the ATF candidate materials, and evaluated fuel behavior simulating operational and accidental conditions by the developed codes. In this paper, the R&D progresses of the ATF candidate materials considered in Japan are reviewed based on the information available such as proceedings of international conference and academic papers, providing an overview of ATF program in Japan.
Li, X.; Sato, Ikken; Yamaji, Akifumi*
Proceedings of 9th Conference on Severe Accident Research (ERMSAR 2019) (Internet), 20 Pages, 2019/03
This study aims at identifying the modeling uncertainties and addressing the sensitivity parameters in Fukushima Daiichi Nuclear Power Station Accident (1F) Unit 3 with MELCOR 2.2 code. Sensitivity studies have been performed on the safety relief valve (SRV) functioning and alternative water injection (AWI) in Unit 3. With the current modelling assumptions in MELCOR, the best-reproduced RPV pressure history of 1F Unit 3 suggested that 6 SRVs should have been open during ADS operation and they remained open when the major core slumping took place at ca. 12:00 on March 13th (ca. 45:20 h after SCRAM). As for lower head failure, there is still large uncertainty in predicting lower head failure time with Larson-Miller creep rupture model in the current MELCOR modeling. The lower head failure timing is not necessarily positively correlated with the amount of water and overall dryout condition of the debris in the lower plenum.
Li, X.; Sato, Ikken; Yamaji, Akifumi*; Duan, G.*
Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 8 Pages, 2018/07
Molten corium-concrete interaction (MCCI) is an important ex-vessel phenomenon that could happen during the late phase of a hypothetical severe accident in a light water reactor. In the present study, a three-dimensional (3-D) numerical study has been performed to simulate COMET-L3 test carried out by KIT with a stratified molten pool configuration of simulant materials with improved MPS method. The heat transfer between corium/crust/concrete was modeled with heat conduction between particles. Moreover, the potential influence of the siliceous aggregates was also investigated by setting up two different case studies since there was previous study indicating that siliceous aggregates in siliceous concrete might contribute to different axial and radial concrete ablation rates. The simulation results have indicated that metal melt as corium in MCCI can have completely different characteristics regarding concrete ablation pattern from that of oxidic corium, which needs to be taken into consideration when assessing the containment melt-through time in severe accident management.
Sakamoto, Kan*; Hirai, Mutsumi*; Ukai, Shigeharu*; Kimura, Akihiko*; Yamaji, Akifumi*; Kusagaya, Kazuyuki*; Kondo, Takao*; Yamashita, Shinichiro
Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 7 Pages, 2017/09
This paper will show the overview of current status of development of accident tolerant FeCrAl-ODS fuel claddings for BWRs (boiling water reactors) in the program sponsored and organized by the Ministry of Economy, Trade and Industry (METI) of Japan. This program is being carried out to create the technical basis for the practical use of the accident tolerant fuels and the other components in LWRs through multifaceted activities. In the development of FeCrAl-ODS fuel claddings both the experimental and the analytical studies have been performed. The acquisition and accumulation of key material properties of FeCrAl-ODS fuel claddings were conducted by using bar, sheet and tube shaped FeCrAl-ODS materials fabricated in this program to support the evaluations in the analytical studies. A neutron irradiation test was also started in the ORNL High Flux Isotope Reactor (HFIR) to examine the effect of neutron irradiation on the mechanical properties.
Yamaji, Akifumi*; Yamasaki, Daiki*; Okada, Tomoya*; Sakamoto, Kan*; Yamashita, Shinichiro
Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 10 Pages, 2017/09
Features of the accident tolerant fuel performance were evaluated with FEMAXI-7 when the current Zircaloy(Zry) cladding is replaced with FeCrAl-ODS steel cladding (a type of oxide dispersion strengthened steel being developed under the Project on Development of Technical Basis for Safety Improvement at Nuclear Power Plants by Ministry of Economy, Trade and Industry (METI) of Japan) for BWR 9
9 type fuel rod. In particular, influences of the creep strain rate and thickness of the ODS cladding on the fuel temperature, fission gas release rate (FGR) and pellet-cladding mechanical interaction (PCMI) are investigated.
Buiron, L.*; Rimpault, G*; Fontaine, B.*; Kim, T. K.*; Stauff, N. E.*; Taiwo, T. A.*; Yamaji, Akifumi*; Gulliford, J.*; Fridmann, E.*; Pataki, I.*; et al.
Proceedings of International Conference on the Physics of Reactors; The Role of Reactor Physics toward a Sustainable Future (PHYSOR 2014) (CD-ROM), 16 Pages, 2014/09
Within the activities of the Working Party on Scientific Issues of Reactor Systems (WPRS) of the OECD, an international collaboration is ongoing on the neutronic analyses of several Generation-IV Sodium-cooled Fast Reactor (SFR) concepts. This paper summarizes the results obtained by participants from institutions of different countries (ANL, CEA, ENEA, HZDR, JAEA, CER, KIT, UIUC) for the large core numerical benchmarks. These results have been obtained using different calculation methods and analysis tools to estimate the core reactivity and isotopic composition evolution, neutronic feedbacks and power distribution. For the different core concepts analyzed, a satisfactory agreement was obtained between participants despite the different calculation schemes used. A good agreement was generally obtained when comparing compositions after burnup, the delayed neutron fraction, the Doppler coefficient, and the sodium void worth. However, some noticeable discrepancies between the k-effective values were observed and are explained in this paper. These are mostly due to the different neutronic libraries employed (JEFF3.1, ENDFB7.0 or JENDL-4.0) and to a lesser extent the calculations methods.
Okajima, Shigeaki; Fougeras, P.*; Gil, C.-S.*; Glinatsis, G.*; Gulliford, J.*; Iwamoto, Osamu; Jacqmin, R.*; Khomyakov, Y.*; Kochetkov, A.*; Kormilitsyn, M. V.*; et al.
NEA/NSC/DOC(2013)3, p.265 - 278, 2013/04
The Expert Group on Integral Experiments for Minor Actinide Management (EG on IEMAM) was established under OECD/NEA/NSC. The objectives are to review integral experiments for validating MA nuclear data, to recommend additional integral experiments and to propose an international framework to facilitate them from view points of the MA management. The paper summarized the discussion results in the EG on IEMAM as follows: (1) Requirement of nuclear data for MA management, (2) Reviewing existing integral data and identifying specification of missing experimental work to be required, (3) Identifying the bottlenecks and considering possible solutions to them and (4) Proposal of action program for international cooperation.
Uematsu, Mari Mariannu; Sugino, Kazuteru; Kawashima, Katsuyuki; Okano, Yasushi; Yamaji, Akifumi; Naganuma, Masayuki; Oki, Shigeo; Okubo, Tsutomu; Ota, Hirokazu*; Ogata, Takanari*; et al.
JAEA-Research 2012-041, 126 Pages, 2013/02
JAEA-Research-2012-041.pdf:16.49MB
The characteristics of sodium-cooled metal fuel core compared to MOX fuel core are given by its higher heavy metal density and superior neutron economy. By taking advantage of these characteristics and allowing flexibility in metal fuel specification and core design conditions as sodium void reactivity and bundle pressure drop, core design with high burnup, high breeding ratio and low fuel inventory features will be achievable. On ground of the major achievements in metal fuels utilization as driver fuels in sodium fast reactors in U.S., the metal fuel core concept is selected as a possible alternative of MOX fuel core concept in FaCT project. This report describes the following items as a result of the joint study on "Reactor core and fuel design of metal fuel core of sodium-cooled fast reactor" conducted by JAEA and CRIEPI during 4 years from fiscal year 2009 to 2012.
Yamaji, Akifumi; Nakano, Yoshihiro; Uchikawa, Sadao; Okubo, Tsutomu
Nuclear Technology, 179(3), p.309 - 322, 2012/09
Times Cited Count：5 Percentile：35.65(Nuclear Science & Technology)HC-FLWR effectively utilizes the uranium (U) and the plutonium (Pu) resources by achieving a fissile Pu conversion ratio of 0.84 without a significant technical gap from the current BWR technology. In this study, a new core design concept for HC-FLWR has been developed to achieve the conversion ratio of 0.95. The concept of the FLWR/MIX fuel assembly, which had been originally proposed for tight fuel bundle, was used to raise the conversion ratio without deteriorating the core void reactivity characteristics. For a semi-tight fuel rod lattice with rod clearance of 0.20 to 0.25 cm, the design ranges of the conversion ratio and the average discharge burnup are 0.91 to 0.94 and 53 to 49 GWd/t, respectively. The conversion ratio can be raised to 0.97 by increasing the 
U enrichment from 4.9 to 6.0 wt%. Two representative core designs and one alternative design option have been obtained. Hence, the flexibility of HC-FLWR concept to achieve the conversion ratio of 0.84 to 0.95 has been revealed.
