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compoundsFrazer, D.*; Saleh, T. A.*; Matsumoto, Taku; Hirooka, Shun; Kato, Masato; McClellan, K.*; White, J. T.*
Nuclear Engineering and Design, 423, p.113136_1 - 113136_7, 2024/07
Nanoindentation based techniques can be employed on minute volumes of material to measure mechanical properties, including Young's modulus, hardness, and creep stress exponents. In this study, (U,Ce)O solid solutions samples are used to develop elevated temperature nanoindentation and nanoindentation creep testing methods for use on mixed oxide fuels. Nanoindentation testing was performed on 3 separate (Ux-1,Cex)O compounds ranging from x equals 0.1 to 0.3 at up to 800 
C: their Young's modulus, hardness, and creep stress exponents were evaluated. The Young's modulus decreases in the expected linear manner while the hardness decreases in the expected exponential manner. The nanoindentation creep experiments at 800 C give stress exponent values, n=4.7-6.9, that suggests dislocation motion as the deformation mechanism.
Sato, Ikken; Yoshikawa, Shinji; Yamashita, Takuya; Shimomura, Kenta; Cibula, M.*; Mizokami, Shinya*
Nuclear Engineering and Design, 422, p.113088_1 - 113088_24, 2024/06
Baccou, J.*; Glantz, T.*; Ghione, A.*; Sargentini, L.*; Fillion, P.*; Damblin, G.*; Sueur, R.*; Iooss, B.*; Fang, J.*; Liu, J.*; et al.
Nuclear Engineering and Design, 421, p.113035_1 - 113035_16, 2024/05
Soma, Shu; Ishigaki, Masahiro*; Abe, Satoshi; Shibamoto, Yasuteru
Nuclear Engineering and Design, 416, p.112754_1 - 112754_18, 2024/01
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Sato, Ikken; Yoshikawa, Shinji; Yamashita, Takuya; Shimomura, Kenta; Cibula, M.*; Mizokami, Shinya*
Nuclear Engineering and Design, 414, p.112574_1 - 112574_20, 2023/12
Uchibori, Akihiro; Doda, Norihiro; Aoyagi, Mitsuhiro; Sonehara, Masateru; Sogabe, Joji; Okano, Yasushi; Takata, Takashi*; Tanaka, Masaaki; Enuma, Yasuhiro; Wakai, Takashi; et al.
Nuclear Engineering and Design, 413, p.112492_1 - 112492_10, 2023/11
Times Cited Count：1 Percentile：72.91(Nuclear Science & Technology)The ARKAIDA has been developed to realize automatic optimization of plant design from safety evaluation for the advanced reactors represented by a sodium-cooled fast reactor. ARKADIA-Design offers functions to support design optimization both in normal operating conditions and design basis events. The multi-level simulation approach by the coupled analysis such as neutronics, core deformation, core thermal hydraulics was developed as one of the main technologies. On the other hand, ARKAIDA-Safety aims for safety evaluation considering severe accidents. As a key technology, the numerical methods for in- and ex-vessel coupled phenomena during severe accidents in sodium-cooled fast reactors were tested through a hypothetical severe accident event. Improvement of the ex-vessel model and development of the AI technology to find best design solution have been started.
Narukawa, Takafumi; Hamaguchi, Shusuke*; Takata, Takashi*; Udagawa, Yutaka
Nuclear Engineering and Design, 411, p.112443_1 - 112443_12, 2023/09
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Kitamura, Akihiro; Hirano, Hiroshi*; Yoshida, Masato
Nuclear Engineering and Design, 411, p.112435_1 - 112435_14, 2023/09
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)This study presents the features and brief history of the glovebox dismantling facility and the primary dismantlement results. Subsequently, we evaluate the novelties of the facility from operational experiences in manual and remote glovebox dismantlement methods and discuss their characteristics. Furthermore, we evaluate the worker exposure dose based on the obtained data. Finally, we show how these experiences are effectively fed back to the technological dismantlement development for our decommissioning project.
Iwasawa, Yuzuru; Sugiyama, Tomoyuki; Kaneko, Akiko*
Nuclear Engineering and Design, 409, p.112348_1 - 112348_15, 2023/08
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Doda, Norihiro; Igawa, Kenichi*; Iwasaki, Takashi*; Murakami, Satoshi*; Tanaka, Masaaki
Nuclear Engineering and Design, 410, p.112377_1 - 112377_15, 2023/08
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)To enhance the safety of sodium-cooled fast reactors, the decay heat in the core must be removed by natural circulation even if the AC power supply to the forced circulation equipment is lost. Under natural circulation conditions, sodium flow is driven by buoyancy, and flow velocity and temperature distribution influence each other. Thus, it is difficult to evaluate the core hot spot temperature by deterministically considering the uncertainties affecting flow and heat. In this study, a statistical evaluation method is developed for the core hot spot temperature by using Monte Carlo sampling methods. The applicability of the core hotspot evaluation method was confirmed in three representative events during natural circulation decay heat removal operations in loop-type sodium-cooled fast reactors.
Abe, Satoshi; Okagaki, Yuria
Nuclear Engineering and Design, 404, p.112165_1 - 112165_14, 2023/04
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Dong, F.*; Chen, S.*; Demachi, Kazuyuki*; Yoshikawa, Masanori; Seki, Akiyuki; Takaya, Shigeru
Nuclear Engineering and Design, 404, p.112161_1 - 112161_15, 2023/04
Times Cited Count：3 Percentile：90.12(Nuclear Science & Technology)Sato, Ikken; Yoshikawa, Shinji; Yamashita, Takuya; Cibula, M.*; Mizokami, Shinya*
Nuclear Engineering and Design, 404, p.112205_1 - 112205_21, 2023/04
Times Cited Count：2 Percentile：90.12(Nuclear Science & Technology)Based on updated knowledge from plant-internal investigations, experiments and model simulations until now, the in-vessel phase of Fukushima-Daiichi Nuclear Power Station Unit 2 was analyzed using the MAAP code. In Unit 2, it is considered that the core material enthalpy was relatively low when it relocated to the lower plenum of the pressure vessel, then, cooled by the coolant and solidified there. Although the MAAP code tended to underestimate the degree of core-material oxidation during the relocation, this probable underestimation was compensated for by an existing study that was considered more reliable, so that more realistic debris conditions in the lower plenum could be obtained. Basic validity of the former prediction of the Unit 2 accident progression behavior was confirmed and detailed boundary condition for the later phase was provided. This boundary condition should be utilized for future studies addressing debris reheating process leading to lower head failure and debris relocation toward the pedestal.
Miyahara, Shinya*; Arita, Yuji*; Nakano, Keita; Maekawa, Fujio; Sasa, Toshinobu; Obayashi, Hironari; Takei, Hayanori
Nuclear Engineering and Design, 403, p.112147_1 - 112147_17, 2023/03
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)It is important to evaluate the inventories and the release and transport behavior of the spallation products (SPs) in the Lead-Bismuth Eutectic (LBE) coolant system of Accelerator Driven System (ADS) for the safety studies of the radiological hazard both in the cases of normal operation and accident. University of Fukui and JAEA have been developing the computer analysis code TRAIL (Transport of RAdionuclides In Liquid metal systems) which predicts the time dependent behavior of SPs within the LBE coolant system of ADS for the wide range of operational events. The source term of both radioactive and stable SPs in the LBE coolant is given as input and the radioactive decay chain model for the radioactive SPs is implemented in the code to evaluate the effect of precursors on the SPs mobility. This paper presents the recent advancement status of the code development and the validation results comparing with the distribution data of volatile SPs in MEGAPIE spallation target.
Yamaguchi, Toshihiko; Mihalache, O.
Nuclear Engineering and Design, 401, p.112084_1 - 112084_14, 2023/01
Times Cited Count：1 Percentile：31.61(Nuclear Science & Technology)Fukaya, Yuji; Okita, Shoichiro; Sasaki, Koei; Ueta, Shohei; Goto, Minoru; Ohashi, Hirofumi; Yan, X.
Nuclear Engineering and Design, 399, p.112033_1 - 112033_9, 2022/12
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Kernel migration of TRi-structural ISOtropic (TRISO) fuel for High Temperature Gas-cooled Reactor (HTGR) has been analyzed to investigate the potential dominating effects. Kernel migration is a major fuel failure mode and dominant to determine the lifetime of the fuel for High Temperature engineering Test Reactor (HTTR). However, this study shows that the result and reliability depend on the evaluation method. The evaluation method used in this study takes into account of actual distribution of Coated Fuel Particles (CFPs) and the resulting heterogeneous fuel temperature calculation with such distribution. The result shows that the Kernel Migration Rate (KMR) is predicted to be about 10% less compared with the most conservative evaluation.
Pacio, J.*; Van Tichelen, K.*; Eckert, S.*; Wondrak, T.*; Di Piazza, I.*; Lorusso, P.*; Tarantino, M.*; Daubner, M.*; Litfin, K.*; Ariyoshi, Gen; et al.
Nuclear Engineering and Design, 399, p.112010_1 - 112010_15, 2022/12
Times Cited Count：5 Percentile：84.97(Nuclear Science & Technology)Heavy-liquid metals (HLMs), such as lead and lead-bismuth eutectic (LBE), are proposed as primary coolants in accelerator driven systems and next-generation fast reactors. In Europe, the reference systems using HLMs are MYRRHA (LBE) and ALFRED (lead). This article presents an overview of recent experiences and ongoing activities on pool-type and loop-type HLM experiments. Pool tests include the measurement of forced- and natural-circulation flow patterns in several scenarios representative of nominal and decay heat removal conditions. Loop tests are focused on the evaluation of specific components, like mockups of the fuel assembly, control rod and heat exchangers. They involve the measurement of global variables, such as flow rate and pressure difference, and local quantities like temperature, velocity and vibrations. Advanced instrumentation, capable of sustaining high temperatures and corrosion, is necessary for accurate measurements, often in compact geometries. In addition to traditional techniques, other instrumentation based on optical fibers, ultrasonic and electromagnetic methods are discussed.
Shibata, Taiju; Nishihara, Tetsuo; Kubo, Shinji; Sato, Hiroyuki; Sakaba, Nariaki; Kunitomi, Kazuhiko
Nuclear Engineering and Design, 398, p.111964_1 - 111964_4, 2022/11
Times Cited Count：2 Percentile：53.91(Nuclear Science & Technology)Japan Atomic Energy Agency (JAEA) has been promoting the research and development (R&D) of High Temperature Gas-cooled Reactor (HTGR). R&D on reactor technologies is carried out by using High Temperature engineering Test Reactor (HTTR). The HTTR was resumed without significant reinforcements in 2021. On January 2022, a safety demonstration test under the OECD/NEA LOFC project was carried out. JAEA is promoting R&D on a carbon-free hydrogen production by thermochemical water splitting Iodine-Sulfur process (IS process). JAEA conducts design study for various HTGR systems toward commercialization. A new test program about demonstration of hydrogen production by the HTTR was launched. Steam methane reforming hydrogen production system was selected for the first demonstration by 2030.
Sotsu, Masutake
Nuclear Engineering and Design, 396( ), p.111893_1 - 111893_27, 2022/09
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Applicability of the analysis model of the cooling ability under a high temperature and low flowrate condition should be improved to evaluate the plant safety in case of the severe accident due to long-term station blackout. The present study reviews experiments on pressure drop behavior for complicated tube bundle geometry to the flow path and then develops a new correlation equation based on a computational fluid dynamics analysis of the Monju air cooler reflecting the actual geometry and plant data. After a basic simulation model being developed for a typical pressure drop experiment, the simulation is applied to the Monju air cooler that has a finned tube bundle. The obtained relationship between friction factors and Reynolds numbers ranging from 10 to 10,000 are fitted to a power function to derive a correlation equation of the fin tube bundle friction factor. The derived correlation equation can estimate pressure loss in the finned tube bundle more precisely than that in the Monju design. It is applicable to the future reactor design of the air cooler, especially when the cooling ability in low Reynolds number is requested.
Ho, H. Q.; Ishii, Toshiaki; Nagasumi, Satoru; Ono, Masato; Shimazaki, Yosuke; Ishitsuka, Etsuo; Goto, Minoru; Simanullang, I. L.*; Fujimoto, Nozomu*; Iigaki, Kazuhiko
Nuclear Engineering and Design, 396, p.111913_1 - 111913_9, 2022/09
Times Cited Count：1 Percentile：31.61(Nuclear Science & Technology)