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Aoyagi, Mitsuhiro; Makino, Toru*; Oki, Hiroshi*; Uchibori, Akihiro; Okano, Yasushi
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 7 Pages, 2023/05
Ohshima, Hiroyuki; Morishita, Masaki*; Aizawa, Kosuke; Ando, Masanori; Ashida, Takashi; Chikazawa, Yoshitaka; Doda, Norihiro; Enuma, Yasuhiro; Ezure, Toshiki; Fukano, Yoshitaka; et al.
Sodium-cooled Fast Reactors; JSME Series in Thermal and Nuclear Power Generation, Vol.3, 631 Pages, 2022/07
This book is a collection of the past experience of design, construction, and operation of two reactors, the latest knowledge and technology for SFR designs, and the future prospects of SFR development in Japan. It is intended to provide the perspective and the relevant knowledge to enable readers to become more familiar with SFR technology.
Yokoyama, Kenji; Maruyama, Shuhei; Taninaka, Hiroshi; Oki, Shigeo
JAEA-Data/Code 2021-019, 115 Pages, 2022/03
In JAEA, several versions of unified cross-section set for fast reactors have been developed so far; we have developed a new unified cross-section set ADJ2017R, which is an improved version of the unified cross-section setADJ2017 for fast reactors. The unified cross-section set is used for reflecting information of C/E values (analysis / experiment values) obtained by integral experiment analyses in reactor core design via the cross-section adjustment methodology; the values are stored in the standard database for FBR core design. In the methodology, the cross-section set is adjusted by integrating the information such as uncertainty (covariance) of nuclear data, uncertainty of integral experiment / analysis, sensitivity of integral experiment with respect to nuclear data. ADJ2017R basically has the same performance as ADJ2017, but we conducted an additional investigation on ADJ2017 and revised the following two points. The first is to unify the evaluation method of the correlation coefficient of uncertainty caused by experiments (hereinafter referred to as the experimental correlation coefficient). Because it was found that the common uncertainty used in the evaluation of the experimental correlation coefficient was evaluated by two different methods, the experimental correlation coefficients were revised for all experimental data, and the evaluation method was unified. The second is the review of the integral experiment data used for the cross-section adjustment calculation. It was found that one of the experimental values of composition ratio after irradiation of the Am-243 sample has a problem in uncertainty evaluation because its experimental uncertainty is extremely small compared to the others. The cross-section adjustment calculation was, therefore, redone by excluding the experimental value. In the creation of ADJ2017, a total of 719 data sets were analyzed and evaluated, and eventually adopted 620 integral experimental data sets. In contrast, a total of 61
Ikeda, Kazumi*; Homma, Yuto*; Moriwaki, Hiroyuki*; Oki, Shigeo
Proceedings of 2014 International Congress on the Advances in Nuclear Power Plants (ICAPP 2014) (CD-ROM), p.1175 - 1183, 2014/04
Tanaka, Masaaki; Fujisaki, Tatsuya*; Oki, Hiroshi*
Nihon Kikai Gakkai Kanto Shibu Ibaraki Koenkai 2013 Koen Rombunshu, p.141 - 142, 2013/09
Numerical simulation method to analyze unsteady flow phenomena in the upper plenum and the hot-leg piping system in Japan Sodium cooled Fast Reactor (JSFR) has been developed. A flow-induced vibration in the hot-leg piping is mainly targeted in this study. A numerical model of the hot-leg piping integrated into a full domain of the upper plenum was constructed. By comparing with the existing experimental results, applicability of numerical model was confirmed.
Ohno, Shuji; Ohshima, Hiroyuki; Tajima, Yuji*; Oki, Hiroshi*
Journal of Power and Energy Systems (Internet), 6(2), p.241 - 254, 2012/06
The authors are initiating systematic verification and validation activity to demonstrate the reliability of numerical simulation tool for sodium fire behavior postulated in a fast reactor plant. The activity is in progress with the main focuses on already developed sodium fire analysis codes SPHINCS and AQUA-SF. The events to be evaluated are sodium spray, pool, or combined fire accidents followed by thermodynamic behaviors. The present paper describes that the "Phenomena Identification and Ranking Table" is developed at first to clarify the important validation points in the codes, and that an "Assessment Matrix" is proposed which summarizes the tests for validating the computational models for important phenomena. Furthermore, the paper shows a practical validation with a separate effect test in which the spray droplet combustion model of both codes predicts the burned amount of a sodium droplet with the error mostly less than 30%.
Ohno, Shuji; Ohshima, Hiroyuki; Tajima, Yuji*; Oki, Hiroshi*
Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 10 Pages, 2011/10
Conceptual design study of the JSFR needs the evaluation of thermodynamic consequences in hypothetical sodium fire accident. The authors are therefore initiating systematic V&V activity for sodium fire evaluation tools. The activity is in progress with the main focuses on already developed sodium fire analysis codes SPHINCS and AQUA-SF. The present paper describes that a preliminary "PIRT" is developed at first, and an "assessment matrix" is proposed which summarizes both separate effect tests (SET) and integral effect tests (IET) for validating the computational models or whole code for important phenomena. Furthermore, the paper shows an individual validation with SET in which the spray droplet combustion model of SPHINCS and AQUA-SF predicts the burned amount of a falling sodium droplet with the error mostly less than 30%.
Ishimori, Kenichiro; Kameo, Yutaka; Matsue, Hideaki; Oki, Yoshiyuki*; Nakashima, Mikio; Takahashi, Kuniaki
Applied Radiation and Isotopes, 69(2), p.506 - 510, 2011/02
Times Cited Count:6 Percentile:44.45(Chemistry, Inorganic & Nuclear)In order to establish a simple and rapid analytical method for C in solidified products made from non-metallic low-level radioactive solid wastes by melting treatment, a radiochemical analysis in combination with alkaline fusion as a sample decomposition method was examined. A simulated solidified product containing C, which was prepared by using nuclear reaction N(n, p)C with thermal neutron irradiation, was analyzed by the present method to compare with a conventional radiochemical analysis using oxidizing combustion. The reproducible and quantitative recovery of C from the simulated solidified product indicates that the present method is superior and more efficient for C analysis in solidified products than the conventional method using oxidizing combustion.
Ohno, Shuji; Oki, Hiroshi*; Sugahara, Akihiro*; Ohshima, Hiroyuki
Journal of Nuclear Science and Technology, 48(2), p.205 - 214, 2011/02
Times Cited Count:11 Percentile:64.25(Nuclear Science & Technology)Validation of a numerical simulation method is carried out for thermal stratification phenomena in the reactor vessel upper plenum of advanced sodium-cooled fast reactors. The study mainly focuses on the fundamental applicability of commercial computational fluid dynamics (CFD) codes as well as an in-house code to the evaluation of thermal stratification behavior including the simulation methods such as spatial mesh distribution and RANS-type turbulence models in the analyses.
Ohno, Shuji; Sugahara, Akihiro*; Oki, Hiroshi*; Ohshima, Hiroyuki
Proceedings of 7th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-7) (CD-ROM), 6 Pages, 2010/11
Parametric analyses were carried out as numerical experiments to clarify the basic characteristics of thermal stratification behavior in upper plenum of fast reactors. The analyses were performed changing hypothetically the analytical conditions of sodium flow rate, sodium temperature, and the size of analyzed upper plenum, by using a commercial CFD code FLUENT Ver. 6.3 and the RNG k- turbulence model. The results provided the suggestion that the averaged sodium ascending velocity in the reactor upper plenum region and the sodium temperature difference before and after the transient initiation would be the dominant factors to determine temperature gradient of thermal stratification interface. Further, it was implied that appropriate spatial mesh arrangement in vertical direction around the stratification interface is significant to obtain the accurate numerical solution of interface temperature gradient.
Tanaka, Masaaki; Murakami, Satoshi*; Oki, Hiroshi*; Ohshima, Hiroyuki
Proceedings of Joint International Conference of 7th Supercomputing in Nuclear Application and 3rd Monte Carlo (SNA + MC 2010) (USB Flash Drive), 6 Pages, 2010/10
Strategy of the numerical estimation method development for the thermal fatigue in JSFR at JAEA is explained from the numerical simulation code development step to the application step for practical problems. Numerical simulation codes prepared for the study and outlines of the verification and validation study are briefly described. Numerical results in a T-junction piping system and typical numerical results around typical control rod channels and the blanket fuel subassemblies simulating the JSFR are shown as examples in recent progress.
Ohno, Shuji; Sugahara, Akihiro*; Oki, Hiroshi*; Ohshima, Hiroyuki
Nihon Kikai Gakkai Rombunshu, B, 76(763), p.451 - 453, 2010/03
Three-dimensional thermal-hydraulic analyses have been carried out for a sodium experiment in a relatively simple axis-symmetric geometry using a commercial CFD code in order to validate simulating methods for thermal stratification behavior in an upper plenum of sodium-cooled fast reactor. Detailed comparison between simulated results and experimental measurement has demonstrated that the code reproduced fairly well the fundamental thermal stratification behaviors such as vertical temperature gradient and upward movement of a stratification interface when utilizing high-order discretization scheme and appropriate mesh size. Furthermore, the investigation has clarified the influence of RANS type turbulence models on phenomena predictability; i.e. the standard - model, the RNG - model and the Reynolds Stress Model.
Ohno, Shuji; Ohshima, Hiroyuki; Sugahara, Akihiro*; Oki, Hiroshi*
Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) (CD-ROM), 12 Pages, 2009/09
Multi-dimensional transient thermal hydraulic analyses for a sodium experiment are carried out in order to validate the applicability of basic simulation methods to a typical thermal stratification behavior in a reactor upper plenum region. Simulation predictability of the behavior is evaluated after the results of an existing experiment are reviewed and important fundamental characteristics of the behavior are quantified. The investigated results clarify that CFD codes provide good prediction for fundamental phenomena of steep temperature gradient and gradual rising behavior of the stratification interface under the condition of adopting appropriate mesh size, higher-order discretization scheme, and the RANS turbulence model.
Ohno, Shuji; Oki, Hiroshi*; Sugahara, Akihiro*; Ohshima, Hiroyuki
Nihon Kikai Gakkai Rombunshu, B, 75(751), p.464 - 465, 2009/03
Three-dimensional thermal-hydraulic analyses of thermal stratification phenomena have been carried out in order to validate simulating ability of a single-phase thermal hydraulic code AQUA and commercial CFD codes STAR-CD and FLUENT. The analyses of thermal stratification water experiments demonstrated that the codes reproduced temperature gradient and upward movement of the stratification interface in the case of utilizing appropriate discretization method and computational mesh arrangement for gravitational direction. No remarkable difference was observed between the calculated results with three codes. It was also shown that three turbulence models of the standard k-e model, the RNG k-e model and the RSM predicted fairly well the fundamental stratification behavior observed in the experiments.
Ohno, Shuji; Oki, Hiroshi*; Sugahara, Akihiro*; Ohshima, Hiroyuki
Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 7 Pages, 2008/11
Validation study of numerical simulation method is in progress for thermal stratification phenomena in a reactor vessel upper plenum of advanced sodium-cooled fast reactors. This paper describes the current status of the study using two kinds of thermal stratification experiments and commercial CFD codes STAR-CD, FLUENT, and an in-house code AQUA.
Ishimori, Kenichiro; Oki, Keiichi; Takaizumi, Hirohide; Kameo, Yutaka; Oki, Yoshiyuki*; Nakashima, Mikio
JAEA-Technology 2007-065, 20 Pages, 2008/01
In order to prepare a reference material which is used for radiochemical analysis of solidified products made from non-metallic miscellaneous low level radioactive solid wastes by melting in Nuclear Science Research Institute, Japan Atomic Energy Agency, the preparation method of the reference material was investigated. Under the optimum melting conditions obtained in this report, the reference material containing Np, Am and Cm as -ray emitting nuclides was successfully prepared. From radiochemical analysis of the reference material, the radioactive concentration of respective nuclides was determined to be 0.1880.001 Bq/g for Np, 0.3680.004 Bq/g for Am, 0.4020.010 Bq/g for Cm.
Oki, Hiroshi*; Okano, Yasushi; Yamaguchi, Akira
JNC TN9520 2004-002, 39 Pages, 2004/06
Chemical equilibrium calculaton proguram for generic multi phase & component system "GENESYS" has been develped. This report descibes instruction of GENESYS program as an operatoin manual, which details (1)Operation instruction, (2)Themo-chemical property of chemical species.
Oki, Hiroshi*; Takata, Takashi; Yamaguchi, Akira
JNC TN9400 2004-023, 70 Pages, 2004/04
The influence of ventilation rate on a sodium combustion pahenomenon is large. And, the longutudinal distance between openings strongly affexts the ventilation rate.
Nishikawa, Hiroyuki*; Souno, T.*; Hattori, M.*; Oki, Y.*; Watanabe, E.*; Oikawa, Masakazu*; Arakawa, Kazuo; Kamiya, Tomihiro
JAERI-Review 2003-033, TIARA Annual Report 2002, p.254 - 256, 2003/11
no abstracts in English
Nishikawa, Hiroyuki*; Souno, T.*; Hattori, M.*; Nishihara, Y.*; Oki, Y.*; Watanabe, E.*; Oikawa, Masakazu*; Kamiya, Tomihiro; Arakawa, Kazuo
Nuclear Instruments and Methods in Physics Research B, 191(1-4), p.342 - 345, 2002/05
Times Cited Count:3 Percentile:24.4(Instruments & Instrumentation)no abstracts in English