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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)Lu, K.; Takamizawa, Hisashi; Katsuyama, Jinya; Li, Y.
International Journal of Pressure Vessels and Piping, 199, p.104706_1 - 104706_13, 2022/10
Times Cited Count：3 Percentile：58.29(Engineering, Multidisciplinary)Iwata, Keiko; Hata, Kuniki; Tobita, Toru; Hirota, Takatoshi*; Takamizawa, Hisashi; Chimi, Yasuhiro; Nishiyama, Yutaka
Proceedings of ASME 2021 Pressure Vessels and Piping Conference (PVP 2021) (Internet), 7 Pages, 2021/07
Hirota, Takatoshi*; Nagoshi, Yasuto*; Hojo, Kiminobu*; Okada, Hiroshi*; Takahashi, Akiyuki*; Katsuyama, Jinya; Ueda, Takashi*; Ogawa, Takuya*; Yashirodai, Kenji*; Ohata, Mitsuru*; et al.
Proceedings of ASME 2021 Pressure Vessels and Piping Conference (PVP 2021) (Internet), 9 Pages, 2021/07
Lu, K.; Katsuyama, Jinya; Li, Y.; Yoshimura, Shinobu*
Journal of Pressure Vessel Technology, 143(2), p.021505_1 - 021505_8, 2021/04
Times Cited Count：1 Percentile：10.06(Engineering, Mechanical)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.
Harada, Masahide; Sakai, Motonobu*; Kurosawa, Takashi*; Haga, Katsuhiro
JPS Conference Proceedings (Internet), 33, p.011098_1 - 011098_5, 2021/03
Materials and Life science experimental Facility (MLF) in J-PARC has a high-intense spallation neutron source and a high-intense muon source induced by 3-GeV and 1MW proton beam. To reduce beam loss, components of proton, neutron and muon beam lines are precisely aligned. However, a settlement of MLF caused by a building construction, installation of heavy shields, earthquakes, consolidation of basement, instrument installation (weight increase), and so on, was assumed in the design stage. Therefore, periodic level measurements in MLF was started from the design stage. A few tens of level measuring points are installed inside and outside MLF. The level measurements were done once per month in the construction stage and per year after operation start. In the construction stage of MLF, it was found that the MLF building largely settled by construction of the building and installation of many shields. And it was also found that The Great East Japan Earthquake on March 11, 2011, caused large settlements of MLF attached buildings. Figure 1 shows fluctuations of measured levels at representative points in each area from December 2011. No critical deterioration (small changes within 
1.0 mm) could be detected. However, a few local settlements are found in the MLF building. In the presentation, periodic level measurement results at MLF will be introduced.
Kinoshita, Norikazu*; Nagaoka, Mika; Nakanishi, Takashi*
Science of the Total Environment, 753, p.142087_1 - 142087_10, 2021/01
Times Cited Count：4 Percentile：26.95(Environmental Sciences)The distribution of the anthropogenic radionuclide americium-241 (
Am), a decay product of Pu discharged from atmospheric tests of nuclear weapons, was investigated to resolve its horizontal and vertical migration in the Tropical East Pacific. We analyzed Am concentrations in seawater samples collected in 2003. On comparing the Am concentrations with the previously determined concentrations of 
Pu in the same samples, the vertical profiles of Am were found to be similar to those of Pu. At some stations, the maximum concentration of Am occurred 100-200 m deeper than that of Pu. The Am/Pu ratios in the North Pacific and South Pacific were comparable to one another, and were typical of the ratio for the Pacific. The Am distribution was influenced by the water mass at depths below 400 m. The Am data support the view there is a current flowing at depths of 400-3000 m from the North Pacific through the equator to the South Pacific. In addition, the Am vertical profile was explained by using a box model that considers the decay of Pu and adsorption and scavenging by suspended particles. The different depths for the maximum concentrations of Am and Pu observed at some stations were well explained by the model and by the distribution of CaCO
particles. The residence time of Am in the Pacific was also estimated by using the model.
and chromia-alumina additive fuels under simulated reactivity-initiated accidents; A Comparative analysis with FEMAXI-8Udagawa, Yutaka; Mihara, Takeshi; Taniguchi, Yoshinori; Kakiuchi, Kazuo; Amaya, Masaki
Annals of Nuclear Energy, 139, p.107268_1 - 107268_9, 2020/05
Times Cited Count：3 Percentile：34.82(Nuclear Science & Technology)Suzuki, Hideya; Tsubata, Yasuhiro; Kurosawa, Tatsuya*; Sagawa, Hiroshi*; Matsumura, Tatsuro
Journal of Nuclear Science and Technology, 54(11), p.1163 - 1167, 2017/11
Times Cited Count：26 Percentile：93.1(Nuclear Science & Technology)A highly practical diamide-type extractant, which is an alkyl diamide amine with 2-ethylhexyl alkyl chains (ADAAM(EH)), was investigated for mutual separation of Am(III) and Cm(III). ADAAM(EH) is a multidentate ligand with one soft N-donor atom and two hard O-donor atoms in its central frame. This tridentate arrangement of donor atoms provides selective binding to Am(III) compared to that with Cm(III) in highly acidic media, resulting in separation factors of up to 5.5. A continuous liquid-liquid extraction and stripping test was conducted using a multistage countercurrent mixer-settler extractor with ADAAM(EH) in n-dodecane. In this test, separation of Am(III) and Cm(III) was achieved with very high yield.
Kibe, Satoshi; Fujisaku, Kazuhiko*; Sakamoto, Atsushi; Sano, Yuichi; Takeuchi, Masayuki; Suzuki, Hideya; Tsubata, Yasuhiro; Matsumura, Tatsuro
JAEA-Research 2016-024, 40 Pages, 2017/02
JAEA-Research-2016-024.pdf:6.73MB
The Japan Atomic Energy Agency has been developing some flowsheets with TDdDGA (N,N,N,Ntetradodecyldiglycolamide) extractant to recover MA (minor actinide) from raffinate. In this study, countercurrent experiments with the improved flowsheet, e.g. the addition of alcohol into the solvent for preventing the precipitation, were performed using miniature centrifugal contactors in order to compare the extraction/stripping behavior of each element with the mixer-settler type. As a result, no entrainments were observed and sufficient phase separation was achieved by centrifugal contactors without any abnormal fluid behavior, such as overflow. The extraction and stripping of Ln(III) which show the similar tendencies as MA could be achieved successfully, especially their stripping proceeded more efficiently in centrifugal contactors. This might be due to the increase in stripping rates by improving the flowsheet and to superior phase separation performance of centrifugal contactors.
Asahina, Daisuke*; Aoyagi, Kazuhei; Kim, K.*; Birkholzer, J.*; Birkholzer, J. T.*; Bolander, J. E.*
Computers and Geotechnics, 81, p.195 - 206, 2017/01
Times Cited Count：33 Percentile：82.31(Computer Science, Interdisciplinary Applications)Kibe, Satoshi; Fujisaku, Kazuhiko*; Ambai, Hiromu; Sakamoto, Atsushi; Sano, Yuichi; Takeuchi, Masayuki; Suzuki, Hideya; Tsubata, Yasuhiro; Matsumura, Tatsuro
JAEA-Research 2015-021, 40 Pages, 2016/02
JAEA-Research-2015-021.pdf:2.3MB
The flowsheet with TDdDGA extractant has been being developed for recovering MA from PUREX raffinate. In the previous study, the yields of MA and other elements in countercurrent extraction/stripping experiments using mixer-settlers were not enough for the target and it would be due to the insufficient phase (aqueous/organic) separation. In this study, we carried out countercurrent experiments with surrogate PUREX raffinate using centrifugal contactors which had superior phase separation ability, and evaluated the extraction/stripping behavior of each element. During the operation, abnormal fluid behavior, such as overflow and entrainment, was not observed, and sufficient phase separation was achieved by centrifugal contactors. Extraction behavior of lanthanides was similar to that in mixer-settlers, but their stripping efficiencies decreased. This would be due to shorter residence time in mixing zone.
Nagase, Fumihisa; Fuketa, Toyoshi
NUREG/CP-0192, p.197 - 230, 2005/10
The Japanese regulatory criterion for a loss-of-coolant-accident (LOCA) is based on a threshold of fuel rod fracture during quenching, which was experimentally determined under simulated LOCA conditions. In order to evaluate the fracture threshold of high burn-up fuel rods, JAERI performs integral thermal shock tests simulating LOCA conditions. The tests have been performed with pre-hydrided, unirradiated claddings and high burn-up fuel claddings irradiated to 39 and 44 GWd/t at a PWR. It was shown that fracture/no-fracture threshold primarily depends on the oxidation amount and that the threshold decreases with increases in hydrogen concentration and axial restraint during the quench. It was also shown that fracture conditions of the tested high burn-up fuel claddings are consistent with the fracture threshold derived from unirradiated claddings with similar hydrogen concentrations.
Ban, Yasutoshi; Asakura, Toshihide; Morita, Yasuji
Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 5 Pages, 2005/10
An idea for controlling Np behavior in the Purex process is that Np(VI) extracted by TBP is selectively reduced to Np(V) by salt-free reagents and separated from U and Pu. Allylhydrazine is expected as a selective Np(VI) reductant from a view point of reduction rates for Np(VI) and Pu(IV). To confirm the applicability of allylhydrazine, a continuous counter-current back-extraction test of Np(VI) has been carried out using a miniature mixer-settler that consists of two steps: U-Pu recovery (3 stages) and Np separation (4 stages). Experimental results show that at least 90% of Np in feed are back-extracted and separated from U and Pu, therefore, it is confirmed that allylhydrazine is expected to be a selective salt-free reductant of Np(VI).
Imaizumi, Hirobumi; Ban, Yasutoshi; Sato, Makoto; Asakura, Toshihide; Morita, Yasuji
JAERI-Research 2005-025, 94 Pages, 2005/09
JAERI-Research-2005-025.pdf:13.61MB
Washing properties of n-butylamine compounds, which are decomposable by burning or electrolysis, for dibutylphosphoric acid (DBP) and Zr in real and simulated degraded solvents have been investigated. In experiments with simulated degraded solvents, basic properties of n-butylamine compounds for washing DBP and Zr were examined and optimum condition was obtained. It was confirmed from the simulated degraded solvent washing tests that 0.25 mol/dm
n-butylamine oxalate of pH 2 could effectively remove Zr from the degraded solvent and the solution of pH 4 was effective for DBP, 95% removal of Zr and DBP were obtained by batch washing. The validity of n-butylamine as a degraded solvent wash reagent was shown by the washing test for real degraded solvent that was performed by continuous counter current flow using a miniature mixer-settler. The present study was carried out as a part of the joint study "Research and Development of Process Elements in an Advanced Aqueous Reprocessing" between Japan Nuclear Cycle Development Institute and JAERI.
Yamaguchi, Isoo*; Suzuki, Shinichi; Sasaki, Yuji; Yamagishi, Isao; Matsumura, Tatsuro; Kimura, Takaumi
JAERI-Tech 2005-037, 56 Pages, 2005/07
JAERI-Tech-2005-037.pdf:2.31MB
For the development of the reprocessing of spent nuclear fuels, the solvent extraction using the mixer-settler equipment is greatly available. This method has the advantages of the treatment of the large mass of materials and continuous operations. In case of the application of the mixer-settler devise, the precise calculation using the distribution ratio of metals in order to determine the metal concentration at each stage is indispensable. This calculation is performed in the development of ARTIST process. The metal concentration in each stage of ARTIST process is calculated by the simulation using excel software equipped with counter-current equations. This method is not taken into consideration of the change of acid concentration, therefore, we developed the new method to calculate the metal concentration even after acidity change. This method can calculate not only the metal concentration at each extraction step but also at each stage of mixer-settler. Using this calculation, we evaluated the optimum condition of solvent extraction in ARTIST process.
Hanawa, Satoshi; Ishihara, Masahiro; Motohashi, Yoshinobu*
Zairyo, 54(2), p.201 - 206, 2005/02
no abstracts in English
Tachibana, Yukio; Nakagawa, Shigeaki; Iyoku, Tatsuo
Nuclear Engineering and Design, 233(1-3), p.103 - 112, 2004/10
Times Cited Count：1 Percentile：9.98(Nuclear Science & Technology)no abstracts in English
Tachibana, Yukio; Nakagawa, Shigeaki; Iyoku, Tatsuo
Elevated Temperature Design and Analysis, Nonlinear Analysis, and Plastic Components, 2004 (PVP-Vol.472), p.39 - 44, 2004/07
The reactor pressure vessel (RPV) of the HTTR is 5.5m in inside diameter, 13.2m in inside height, and 122mm and 160mm in wall thickness of the body and the top head dome, respectively. Because the reactor inlet temperature of the HTTR is higher than that of LWRs, 2 1/4Cr-1Mo steel is chosen for the RPV material. Fluence of the RPV is estimated to be less than 1
10
n/cm
(E
1 MeV), and so irradiation embrittlement is presumed to be negligible, but temper embrittlement is not. For the purpose of reducing embrittlement, content of some elements is limited on 2 1/4 Cr-1 Mo steel for the RPV using embrittlement parameters, J-factor and X-bar. In this paper design, fabrication procedure, and in-service inspection technique of the RPV for the HTTR are described.
