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removal from aqueous solutions; Cost-effectiveness & parametric effectsMaamoun, I.; Eljamal, R.*; Eljamal, O.*
Chemosphere, 312, Part 1, p.137176_1 - 137176_11, 2023/01
Times Cited Count：3 Percentile：38.08(Environmental Sciences)Saito, Kimiaki; Takahara, Shogo; Uezu, Yasuhiro
Nihon Genshiryoku Gakkai-Shi ATOMO
, 60(2), p.111 - 115, 2018/02
no abstracts in English
Chimi, Yasuhiro; Takamizawa, Hisashi; Kasahara, Shigeki*; Iwata, Keiko; Nishiyama, Yutaka
Nuclear Engineering and Design, 307, p.411 - 417, 2016/10
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)To investigate influential parameters for irradiation-assisted stress corrosion cracking (IASCC) growth behavior, we attempt to analyze statistically existing data on the crack growth rate (CGR) in irradiated austenitic stainless steels (SSs) in boiling water reactor (BWR) environments using the Bayesian nonparametric (BNP) method. From the probability distribution of CGR and some input parameters, such as yield stress of irradiated material (
), stress intensity factor (
), electrochemical corrosion potential (ECP), and fast neutron fluence, the mean CGR is estimated and compared with the measured CGR. The analytical results show good reproducibility of the measured CGR. The results also indicate the possible neutron fluence effects on CGR in high CGR region (i.e., high neutron fluence condition) by radiation-induced segregation (RIS), localized deformation, and/or other mechanisms than radiation hardening.
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：72.62(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.
; ; Takasaki, Koichi*; ;
JAERI-Review 98-005, 203 Pages, 1998/03
JAERI-Review-98-005.pdf:8.96MB
no abstracts in English
; ; ; ; ;
Nuclear Safeguards Technology 1986,Vol.1, p.341 - 352, 1987/00
no abstracts in English
J.Lovett*; ; J.Shipley*; D.Sellinschegg*
J.Inst.Nucl.Mater.Manage., 12, p.31 - 34, 1983/00
no abstracts in English
J.Lovett*; ; J.Shipley*; D.Selliuschegg*
Nuclear Safeguards Technology,Vol.1, p.487 - 497, 1982/00
no abstracts in English
; ; ; ; ;
JAERI-M 9665, 143 Pages, 1981/10
no abstracts in English
J.E.Lovett*; ; ; R.H.Augustson*
JAERI-M 9186, 46 Pages, 1980/11
no abstracts in English
; ;
JAERI-M 8757, 41 Pages, 1980/03
no abstracts in English
; Osamu Tanaka*
Journal of the Physical Society of Japan, 18(5), p.610 - 619, 1963/00
Times Cited Count：13no abstracts in English
Ueki, Taro
no journal, ,
A new methodology has been developed to make the reliable estimation of statistical errors in Monte Carlo criticality calculation (MCCC). The methodology developed is directly based on the convergence process in the functional central limit theorem and is shown to perform well in the evaluation of reactor power distribution. The theoretical backbones are described within the general context as framed in the operations research. The requisite basics of statistics are reviewed in terms of output analysis in MCCC. Numerical results are presented for the initial core model of a 1200 MWe pressurized water reactor. Preliminary results of fractal dimension analysis are shown to discuss a potential for convergence assessment.
