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Nb by activation method and half-life of 
Nb by mass analysisNakamura, Shoji; Shibahara, Yuji*; Endo, Shunsuke; Kimura, Atsushi
Journal of Nuclear Science and Technology, 60(11), p.1361 - 1371, 2023/11
Times Cited Count：1 Percentile：72.91(Nuclear Science & Technology)The thermal-neutron capture cross section (
) and resonance integral (I) for Nb among nuclides for decommissioning were measured by an activation method and the half-life of Nb by mass analysis. Niobium-93 samples were irradiated with a hydraulic conveyer installed in the research reactor in Institute for Integral Radiation and Nuclear Science, Kyoto University. Gold-aluminum, cobalt-aluminum alloy wires were used to monitor thermal-neutron fluxes and epi-thermal Westcott's indexes at an irradiation position. A 25-
m-thick gadolinium foil was used to sort out reactions ascribe to thermal-and epi-thermal neutrons. Its thickness provided a cut-off energy of 0.133 eV. In order to attenuate radioactivity of 
Ta due to impurities, the Nb samples were cooled for nearly 2 years. The induced radio activity in the monitors and Nb samples were measured by 
-ray spectroscopy. In analysis based on Westcott's convention, the and I values were derived as 1.11
0.04 barn and 10.50.6 barn, respectively. After the -ray measurements, mass analysis was applied to the Nb sample to obtain the reaction rate. By combining data obtained by both -ray spectroscopy and mass analysis, the half-life of Nb was derived as (2.000.15)
10
years.
Yamazaki, Yasuhiro*; Shinomiya, Keisuke*; Okumura, Tadaharu*; Suzuki, Kenji*; Shobu, Takahisa; Nakamura, Yuiga*
Quantum Beam Science (Internet), 7(2), p.14_1 - 14_12, 2023/05
Doi, Daisuke
Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 7 Pages, 2022/08
Kato, Takuma*; Nagaoka, Mika; Guo, H.*; Fujita, Hiroki; Aida, Taku*; Smith, R. L. Jr.*
Environmental Science and Pollution Research, 28(39), p.55725 - 55735, 2021/10
Times Cited Count：0 Percentile：0(Environmental Sciences)In this work, hydrothermal leaching was applied to simulated soils (clay minerals vermiculite, montmorillonite, kaolinite) and actual soils (Terunuma, Japan) to generate organic acids with the objective to develop an additive-free screening method for determination of Sr in soil. Stable strontium (SrCl
) was adsorbed onto soils for study and ten organic acids were evaluated for leaching Sr from simulated soils under hydrothermal conditions (120 to 200
C) at concentrations up to 0.3 M. For strontium-adsorbed vermiculite (Sr-V), 0.1 M citric acid was found to be effective for leaching Sr at 150C and 1 h treatment time. Based on these results, the formation of organic acids from organic matter in Terunuma soil was studied. Hydrothermal treatment of Terunuma soil produced a maximum amount of organic acids at 200C and 0.5 h reaction time. To confirm the possibility for leaching of Sr from Terunuma soil, strontium-adsorbed Terunuma soil (Sr-S) was studied. For Sr-S, hydrothermal treatment at 200C for 0.5 h reaction time allowed 40% of the Sr to be leached at room temperature, thus demonstrating an additive-free method for screening of Sr in soil. The additive-free hydrothermal leaching method avoids calcination of solids in the first step of chemical analysis and has application to both routine monitoring of metals in soils and to emergency situations.
Lu, K.; Katsuyama, Jinya; Li, Y.
Journal of Pressure Vessel Technology, 142(5), p.051501_1 - 051501_10, 2020/10
Times Cited Count：2 Percentile：15.81(Engineering, Mechanical)Nakayama, Masashi
JAEA-Data/Code 2020-009, 98 Pages, 2020/09
JAEA-Data-Code-2020-009.pdf:6.06MB
JAEA-Data-Code-2020-009-appendix1(DVD-ROM).zip:172.11MBJAEA-Data-Code-2020-009-appendix2-1(DVD-ROM).zip:468.85MBJAEA-Data-Code-2020-009-appendix2-2(DVD-ROM).zip:102.94MBJAEA-Data-Code-2020-009-appendix2-3(DVD-ROM).zip:141.88MBJAEA-Data-Code-2020-009-appendix2-4(DVD-ROM).zip:141.6MB
Japan Atomic Energy Agency (JAEA) has developed a low alkaline cement, named as HFSC (Highly Fly-ash contained Silicafume Cement), in order to decrease the effect on geological environment of cementitious material. HFSC was used experimentally as the shotcrete (140m, 250m and 350m depth gallery) and lining concrete (GL-374m to GL-380m of east access shaft) in construction part of Horonobe URL. In order to evaluate the effect of HFSC on the surrounding rock and groundwater, concrete and rock cores were periodically sampled from the site where the in-situ construction test was conducted, and various analyzes were conducted. Ordinary Portland Cement (OPC) was used for part of 140m depth gallery, and the same analysis as HFSC was conducted, in order to compare the effect of OPC and HFSC. This report summarizes the results of analyzes conducted on core samples from 2009 to 2018.
Rizaal, M.; Nakajima, Kunihisa; Saito, Takumi*; Osaka, Masahiko; Okamoto, Koji*
Journal of Nuclear Science and Technology, 57(9), p.1062 - 1073, 2020/09
Times Cited Count：8 Percentile：71.58(Nuclear Science & Technology)The interaction of cesium hydroxide and a calcium silicate insulation material was experimentally investigated at high temperature conditions. A thermogravimetry equipped with differential thermal analysis was used to analyze thermal events in the samples of mixed calcium silicate and cesium hydroxide under Ar-5%H and Ar-4%H-20%H0 with maximum temperature of 1100C. Prior being mixed with cesium hydroxide, a part of calcium silicate was pretreated at high temperature to evaluate the effect of possible structural changes of this material due to a preceding thermal history and also the sake of thermodynamic evaluation to those available ones. Based upon the initial condition (preliminary heat treatment) of calcium silicate, it was found that if the original material consisted of xonotlite (Ca
Si0
(0H)), the endothermic reaction with cesium hydroxide occurred over the temperature range 575-730C meanwhile if the crystal phase of original material was changed to wollastonite (CaSi0
), the interaction occurred over temperature range 700-1100C. Furthermore, the X-ray diffraction analyses have indicated on both type of pretreated calsils that regardless of Ar-5%H and Ar-4%H-20%H0 atmosphere, cesium aluminum silicate, CsAlSi0
was formed with aluminum in the samples as an impurity or adduct.
Herranz, L. E.*; Jacquemain, D.*; Nitheanandan, T.*; Sandberg, N.*; Barr
, F.*; Bechta, S.*; Choi, K.-Y.*; D'Auria, F.*; Lee, R.*; Nakamura, Hideo
Progress in Nuclear Energy, 127, p.103432_1 - 103432_14, 2020/09
Times Cited Count：3 Percentile：11.26(Nuclear Science & Technology)Wan, T.; Obayashi, Hironari; Sasa, Toshinobu
Nuclear Technology, 205(1-2), p.188 - 199, 2019/01
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Nakamura, Hideo
Proceedings of ANS International Conference on Best Estimate Plus Uncertainties Methods (BEPU 2018) (USB Flash Drive), 8 Pages, 2018/00
no abstracts in English
Wan, T.; Obayashi, Hironari; Sasa, Toshinobu
Proceedings of 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-17) (USB Flash Drive), 13 Pages, 2017/09
Akimoto, Hajime; Sugawara, Takanori
JAEA-Data/Code 2016-008, 87 Pages, 2016/09
JAEA-Data-Code-2016-008.pdf:15.62MB
Thermal hydraulic behavior in a lead-bismuth cooled accelerator driven system (ADS) is analyzed under normal operation condition. Input data for the ADS version of J-TRAC code have been constructed to integrate the conceptual design. The core part of the ADS is modeled in detail to evaluate the core radial power profile effect on the core cooling. As the result of the analyses, the followings are found; (1) Both maximum clad temperature and fuel temperature are below the design limits. (2) The radial power profile has little effect on the coolant flow distribution among fuel assemblies. (3) The radial power profile has little effect on the heat transfer coefficients along fuel rods. (4) The thermal hydraulic behaviors along four steam generators are identical. The thermal hydraulic behaviors along two pumps are also identical. A fast running input data is developed by the simplification of the detailed input data based on the findings mentioned above.
Tochio, Daisuke; Fujimoto, Nozomu
Journal of Nuclear Science and Technology, 53(3), p.425 - 431, 2016/03
Times Cited Count：1 Percentile：10.71(Nuclear Science & Technology)The future HTGR is now designed in JAEA. The reactor has many merging points of helium gas with different temperature. It is needed to clear the mixing characteristics of helium gas at the pipe in the HTGR from the viewpoint of structure integrity and temperature control. Previously, the reactor inlet coolant temperature was controlled lower than specific one in the HTTR due to lack of mixing of helium gas in the primary cooling system. Now the control system is improved to use the calculated bulk temperature of reactor inlet helium gas. In this paper, thermal-hydraulic analysis on the primary cooling system of the HTTR was conducted to clarify the mixing behavior of helium gas. As the result, it was confirmed that the mixing behavior of helium gas in the primary cooling system is mainly affected by the aspect ratio of annular flow path, and it is needed to consider the mixing characteristics of helium gas at the piping design of the HTGR.
Watanabe, Osamu*; Oyama, Kazuhiro*; Endo, Junji*; Doda, Norihiro; Ono, Ayako; Kamide, Hideki; Murakami, Takahiro*; Eguchi, Yuzuru*
Journal of Nuclear Science and Technology, 52(9), p.1102 - 1121, 2015/09
Times Cited Count：13 Percentile：73.22(Nuclear Science & Technology)A natural circulation (NC) evaluation methodology has been developed to ensure the safety of a sodium-cooled fast reactor (SFR) of 1500MW adopting the NC decay heat removal system (DHRS). The methodology consists of a 1D safety analysis which can evaluate the core hot spot temperature taking into account the temperature flattening effect in the core, a 3D fluid flow analysis which can evaluate the thermal-hydraulics for local convections and thermal stratifications in the primary system and DHRS, and a statistical safety evaluation method. The safety analysis method and the 3D analysis method have been validated using results of a 1/10 scaled water test simulating the primary system of the SFR and a 1/7 scaled sodium test simulating the primary system and the DHRS, and the applicability of the safety analysis for the SFR has been confirmed by comparing with the 3D analysis. Finally, a statistical safety evaluation has been performed for the SFR using the safety analysis method.
Akimoto, Hajime
JAEA-Data/Code 2014-031, 75 Pages, 2015/03
JAEA-Data-Code-2014-031.pdf:37.23MB
A thermal-hydraulic analysis code for transmutation system with lead-bismuth cooled accelerator-driven system (ADS) has been developed using the Japanese-version of Transient Reactor Analysis Code (J-TRAC) as the framework to apply the design studies of ADS. To identify the required capabilities of the thermal-hydraulic analysis code for ADS, previous thermal-hydraulic analyses of light water reactors, sodium-cooled fast reactor and ADS have been surveyed. To make up for insufficient capabilities of the J-TRAC code as a thermal-hydraulic analysis code of ADS, physical properties of lead-bismuth eutectic (LBE), argon gas and nitride nuclear fuel were implemented to the J-TRAC code. It was confirmed that the implemented capabilities worked as expected through verification calculations on (1) single-phase LBE flow, (2) heat transfer in a fuel assembly, and (3) heat transfer in a steam generator.
Nishihara, Tetsuo; Inagaki, Yoshiyuki
Nuclear Technology, 153(1), p.100 - 106, 2006/01
Times Cited Count：9 Percentile：53.38(Nuclear Science & Technology)Japan Atomic Energy Research Institute (JAERI) has performed the research and development of hydrogen production using the high temperature engineering test reactor (HTTR). One of the key issues for the HTTR hydrogen production system is the development of control technology for stable operation. A thermal load absorber concept using a steam generator installed downstream of a reformer is proposed to mitigate a variation of helium temperature. Thermal hydraulic analyses for the start up operation and the suspension of feed gas supply to the reformer are carried out. These results show that a large variation of the reformer outlet helium temperature takes place due to a change of the feed gas flow rate. However the steam generator can mitigate the variation of helium temperature. It is clarified that the HTTR can continue normal operation independently of the feed gas flow rate.
Takase, Kazuyuki; Yoshida, Hiroyuki; Ose, Yasuo*; Akimoto, Hajime
Computational Fluid Dynamics 2004, p.649 - 654, 2006/00
no abstracts in English
Oigawa, Hiroyuki; Tsujimoto, Kazufumi; Kikuchi, Kenji; Kurata, Yuji; Sasa, Toshinobu; Umeno, Makoto*; Nishihara, Kenji; Saito, Shigeru; Mizumoto, Motoharu; Takano, Hideki*; et al.
Proceedings of 4th International Workshop on the Utilisation and Reliability of High Power Proton Accelerators, p.325 - 334, 2005/11
The Japan Atomic Energy Research Institute (JAERI) is conducting the research and development (R&D) on the Accelerator-Driven Subcritical System (ADS) for the effective transmutation of minor actinides (MAs). The ADS proposed by JAERI is the 800 MWth, Pb-Bi cooled, tank-type subcritical reactor loaded with (MA+Pu) nitride fuel. The Pb-Bi is also used as the spallation target. In this study, the feasibility of the ADS was discussed with putting the focus on the design around the beam window. The partition wall was placed between the target region and the ductless-type fuel assemblies to keep the good cooling performance for the hot-spot fuel pin. The flow control nozzle was installed to cool the beam window effectively. The thermal-hydraulic analysis showed that the maximum temperature at the outer surface of the beam window could be repressed below 500 C even in the case of the maximum beam power of 30 MW. The stress caused by the external pressure and the temperature distribution of the beam window was also below the allowable limit.
Okumura, Keisuke; Kawasaki, Kenji*; Mori, Takamasa
JAERI-Research 2005-018, 64 Pages, 2005/08
JAERI-Research-2005-018.pdf:3.26MB
In the KRITZ-2 critical experiments, criticality and pin power distributions were measured at room temperature and high temperature (about 245 degree C) for three different cores loading slightly enriched UO or MOX fuels. For nuclear data testing, benchmark analysis was carried out with a continuous-energy Monte Carlo code MVP and its four nuclear data libraries based on JENDL-3.2, JENDL-3.3, JEF-2.2 and ENDF/B-VI.8. As a result, fairly good agreements with the experimental data were obtained with any libraries for the pin power distributions. However, the JENDL-3.3 and ENDF/B-VI.8 give under-prediction of criticality and too negative isothermal temperature coefficients for slightly enriched UO cores, while the older nuclear data JENDL-3.2 and JEF-2.2 give rather good agreements with the experimental data. From the detailed study with an infinite unit cell model, it was found that the differences among the libraries are mainly due to the different fission cross section of U-235 in the energy rage below 1.0 eV.
Sumita, Junya; Shibata, Taiju; Nakagawa, Shigeaki; Hanawa, Satoshi; Iyoku, Tatsuo; Ishihara, Masahiro
Transactions of 18th International Conference on Structural Mechanics in Reactor Technology (SMiRT-18), p.4822 - 4828, 2005/08
Graphite materials are used for structural components in High Temperature Gas-Cooled Reactor (HTGR) core because of their excellent thermo/mechanical properties. Thermal conductivity of graphite components is reduced by neutron irradiation in reactor operation. The reduced conductivity is expected to be recovered by thermal annealing effect when irradiated graphite component is heated above irradiated temperature. In the present study, temperature analyses considering the annealing effect of the HTGR core at a depressurization accident were carried out and influence of annealing effect on maximum fuel temperature was investigated. The analyses show that the annealing effect can reduce the fuel temperature about 100C at the maximum, and it is possible to evaluate the maximum fuel temperature more appropriately. It was also shown that the core-temperature of High Temperature Engineering Test Reactor (HTTR) at the safety demonstration tests can be analyzed with the developed evaluation method considering annealing effect.
