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Soma, Shu; Ishigaki, Masahiro*; Abe, Satoshi; Shibamoto, Yasuteru
Nuclear Engineering and Technology, 56(7), p.2524 - 2533, 2024/07
Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)Tagami, Hirotaka; Tobita, Yoshiharu
Nuclear Engineering and Technology, 56(3), p.873 - 879, 2024/03
Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)A sodium-cooled fast reactor (SFR) core has a potential of prompt criticality due to a change of core material distribution during a severe accident, and the resultant energy release has been one of the safety issues of SFRs. In this study, the safety assessment of an unprotected loss-of-flow (ULOF) in a small SFR (SSFR) has been performed using the SIMMER-IV computer code, which couples the models of space- and time-dependent neutronics and multi-component, multi-field thermal hydraulics in three dimensions. The code, therefore, is applicable to the simulations of transient behaviors of extended disrupted core material motion and its reactivity effects during the transition phase (TP) of ULOF, including a potential of prompt-criticality power excursions driven by fuel compaction. Several conservative assumptions are used in the TP analysis by SIMMER-IV. It was found out that one of the important mechanisms that drives the reactivity-inserting fuel motion was sodium vapor pressure resulted from a fuel-coolant interaction (FCI), which itself was non-energetic local phenomenon. The uncertainties relating to FCI is also evaluated in much conservative way in the sensitivity analysis. From this study, the ULOF characteristics in an SSFR have been understood. Occurrence of recriticality events under conservative assumptions are plausible, but their energy releases are limited.
Yamamoto, Tomohiko; Kato, Atsushi; Hayakawa, Masato; Shimoyama, Kazuhito; Ara, Kuniaki; Hatakeyama, Nozomu*; Yamauchi, Kanau*; Eda, Yuhei*; Yui, Masahiro*
Nuclear Engineering and Technology, 56(3), p.893 - 899, 2024/03
Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)Osawa, Naoki*; Kim, S.-Y.*; Kubota, Masahiko*; Wu, H.*; Watanabe, So; Ito, Tatsuya; Nagaishi, Ryuji
Nuclear Engineering and Technology, 56(3), p.812 - 818, 2024/03
Times Cited Count:1 Percentile:75.38(Nuclear Science & Technology)Zablackaite, G.; Shiotsu, Hiroyuki; Kido, Kentaro; Sugiyama, Tomoyuki
Nuclear Engineering and Technology, 56(2), p.536 - 545, 2024/02
Times Cited Count:1 Percentile:75.38(Nuclear Science & Technology)Furutaka, Kazuyoshi; Ozu, Akira; Toh, Yosuke
Nuclear Engineering and Technology, 55(11), p.4002 - 4018, 2023/11
Times Cited Count:1 Percentile:34.39(Nuclear Science & Technology)Abe, Satoshi; Shibamoto, Yasuteru
Nuclear Engineering and Technology, 55(5), p.1742 - 1756, 2023/05
Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)Tonna, Ryutaro*; Sasaki, Takayuki*; Kodama, Yuji*; Kobayashi, Taishi*; Akiyama, Daisuke*; Kirishima, Akira*; Sato, Nobuaki*; Kumagai, Yuta; Kusaka, Ryoji; Watanabe, Masayuki
Nuclear Engineering and Technology, 55(4), p.1300 - 1309, 2023/04
Times Cited Count:4 Percentile:82.11(Nuclear Science & Technology)Simulated debris was synthesized using UO, Zr, and stainless steel and a heat treatment method under inert or oxidizing conditions. The primary U solid phase of the debris synthesized at 1473 K under inert conditions was UO
, whereas a (U,Zr)O
solid solution formed at 1873 K. Under oxidizing conditions, a mixture of U
O
and (Fe,Cr)UO
phases formed at 1473 K whereas a (U,Zr)O
solid solution formed at 1873 K. The leaching behavior of the fission products from the simulated debris was evaluated using two methods: the irradiation method, for which fission products were produced via neutron irradiation, and the doping method, for which trace amounts of non-radioactive elements were doped into the debris. The dissolution behavior of U depended on the properties of the debris and aqueous medium the debris was immersed in. Cs, Sr, and Ba leached out regardless of the primary solid phases. The leaching of high-valence Eu and Ru ions was suppressed, possibly owing to their solid-solution reaction with or incorporation into the uranium compounds of the simulated debris.
Fukai, Hirofumi*; Furuya, Masahiro*; Yamano, Hidemasa
Nuclear Engineering and Technology, 55(3), p.902 - 907, 2023/03
Times Cited Count:4 Percentile:82.11(Nuclear Science & Technology)This paper addresses reaction products and their distribution of the eutectic melting/solidifying reaction of boron carbide (BC) and stainless-steel (SS). The influence of the existence of carbon on the B
C-SS eutectic reaction was investigated by comparing the iron boride (FeB)-SS reaction by Raman spectroscopy with Multivariate Curve Resolution (MCR) analysis. The scanning electron microscopy with dispersive X-ray spectrometer was also used to investigate the elemental information of the pure metals such as Cr, Ni, and Fe. In the B
C-SS samples, a new layer was formed between B
C/SS interface, and the layer was confirmed that the formed layer corresponded to amorphous carbon (graphite) or FeB or Fe
B. In contrast, a new layer was not clearly formed between FeB and SS interface in the FeB-SS samples.
Shaimerdenov, A.*; Gizatulin, S.*; Dyussambayev, D.*; Askerbekov, S.*; Ueta, Shohei; Aihara, Jun; Shibata, Taiju; Sakaba, Nariaki
Nuclear Engineering and Technology, 54(8), p.2792 - 2800, 2022/08
Times Cited Count:8 Percentile:79.63(Nuclear Science & Technology)Kawada, Kenichi; Suzuki, Toru*
Nuclear Engineering and Technology, 53(12), p.3930 - 3943, 2021/12
Times Cited Count:1 Percentile:11.39(Nuclear Science & Technology)To improve the capability of the SAS4A code, which simulates the initiating phase of core disruptive accidents for MOX-fueled Sodium-cooled Fast Reactors (SFRs), the authors have investigated in detail the physical phenomena under unprotected loss-of-flow (ULOF) conditions in a previous paper. As the conclusion of the last article, fuel stub motion, in which the residual fuel pellets would move toward the core central region after fuel pin disruption, was identified as one of the key phenomena to be appropriately simulated for the initiating phase of ULOF. In the present paper, a simple model describing fuel stub motion, which was not modeled in the previous SAS4A code, was newly proposed. The applicability of the proposed model was validated through a series of analyses for the CABRI experiments, by which the stub motion would be represented with reasonable conservativeness for the reactivity evaluation of disrupted core.
Takeda, Takeshi; Otsu, Iwao
Nuclear Engineering and Technology, 50(6), p.829 - 841, 2018/08
Times Cited Count:14 Percentile:77.71(Nuclear Science & Technology)Takeda, Takeshi; Otsu, Iwao
Nuclear Engineering and Technology, 49(5), p.928 - 940, 2017/08
Times Cited Count:5 Percentile:40.90(Nuclear Science & Technology)Zheng, X.; Ishikawa, Jun; Sugiyama, Tomoyuki; Maruyama, Yu
Nuclear Engineering and Technology, 49(2), p.434 - 441, 2017/03
Times Cited Count:5 Percentile:40.90(Nuclear Science & Technology)Suzuki, Toru; Tobita, Yoshiharu; Kawada, Kenichi; Tagami, Hirotaka; Sogabe, Joji; Matsuba, Kenichi; Ito, Kei; Ohshima, Hiroyuki
Nuclear Engineering and Technology, 47(3), p.240 - 252, 2015/04
Times Cited Count:28 Percentile:90.34(Nuclear Science & Technology)Cheng, S.; Yamano, Hidemasa; Suzuki, Toru; Tobita, Yoshiharu; Nakamura, Yuya*; Zhang, B.*; Matsumoto, Tatsuya*; Morita, Koji*
Nuclear Engineering and Technology, 45(3), p.323 - 334, 2013/06
Times Cited Count:39 Percentile:93.50(Nuclear Science & Technology)Sawada, Atsushi; Sato, Hisashi
Nuclear Engineering and Technology, 42(1), p.9 - 16, 2010/02
Experimental studies for evaluating fracture were conducted by using transparent replica of a single fracture, in order to obtain data for contributing to the methodology improving how we define representative parameter values used for a parallel plate fracture model. Quantitative aperture distribution and tracer concentration data were obtained by measuring the attenuation of transmitted light through the fracture, in high spatial resolution. The representative aperture values evaluated from the multiple measurement methods, such as arithmetic mean of aperture distribution, transport aperture, average aperture evaluated from fracture void volume measurement, converged to a unique value, which indicates the accuracy of this experimental study. The aperture data was employed for studying the validity of numerical simulation under the assumption of local cubic law and showed that the calculated flow rate through the fracture is 10% - 100% larger than hydraulic test results.
Smith, G.*; Kato, Tomoko
Nuclear Engineering and Technology, 42(1), p.1 - 8, 2010/02
Geological disposal is designed to provide safe containment of radioactive waste for very long times, with the containment provided by a combination of engineered and geological barriers. In the extreme long term, after many thousands of years or longer, residual amounts of long-lived radionulides such as Cl-36, but also radionuclides in the natural decay chains, may be released into the environment normally accessed and used by humans, termed here, the biosphere. It is necessary to ensure that any such releases meet radiation protection objectives through the development of a safety case, which will include assessment of radiation doses to humans. The design of such dose calculations over such long timeframes is not straightforward, because of the range of potentially relevant assumptions which could be made, concerning environmental change and changes in human behavior. The issue has therefore been subject to international cooperation for many years. This paper summarizes the evolution and results of that collaboration leading up to the present day, taking account of developments in international recommendations on radiation protection objectives and the more recent greater focus on preparation of site specific safety cases.
Chuto, Toshinori; Nagase, Fumihisa; Fuketa, Toyoshi
Nuclear Engineering and Technology, 41(2), p.163 - 170, 2009/03
In order to evaluate high-temperature oxidation behavior of the advanced alloy cladding under LOCA conditions, isothermal oxidation tests in steam were performed with cladding specimens prepared from high burnup PWR fuel rods (79 MWd/kg). Cladding materials were M5 and ZIRLO which are Nb-containing alloys. Ring-shaped specimens were isothermally oxidized in flowing steam at temperatures from 1173 to 1473 K for the duration between 120 and 4000 s. Protective effect of the pre-formed corrosion layer is seen for the shorter time range at the lower temperatures. Influence of pre-hydriding is not significant for the examined range. Alloy composition change generally has small influence on the oxidation in the examined temperature range, though M5 shows obviously smaller oxidation constant at 1273 K. Consequently, the oxidation rates of the high burnup M5 and ZIRLO cladding are comparable or lower than that of unirradiated Zircaloy-4 cladding.
Kawata, Tomio
Nuclear Engineering and Technology, 39(6), p.683 - 690, 2007/12
To make nuclear energy truly sustainable, it is necessary to ensure not only the sustainability of the fuel supply but also the sustained availability of waste repositories, especially those for high-level radioactive waste (HLW). From this perspective, the effort to maximize the waste loading density in a given repository is important for easing repository capacity problems. In most cases, the loading of a repository is controlled by the decay heat of the emplaced waste. In this paper, a comparison of the decay heat characteristics of HLW is made among the various fuel cycle options. It is suggested that, for a future fast breeder reactor (FBR) cycle, the removal and burning of minor actinides (MA) would significantly reduce the heat load in waste and would allow for a reduction of repository size by half.