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Tanaka, Kosuke; Sato, Isamu*; Onishi, Takashi; Ishikawa, Takashi; Hirosawa, Takashi; Katsuyama, Kozo; Seino, Hiroshi; Ohno, Shuji; Hamada, Hirotsugu; Tokoro, Daishiro*; et al.
Journal of Nuclear Materials, 536, p.152119_1 - 152119_8, 2020/08
Times Cited Count：0 Percentile：0.01(Materials Science, Multidisciplinary)In order to obtain the release rate coefficients from fuels for fast reactors (FRs), heating tests and the subsequent analyses of the fission products (FPs) and actinides that are released were carried out using samples of uranium-plutonium mixed oxide (MOX) fuel pellets irradiated at the experimental fast reactor Joyo. Three heating tests targeting temperatures of 2773, 2973 and 3173 K were conducted using an FP release behavior test apparatus equipped with a high-frequency induction furnace and solid FP sampling systems consisting of a thermal gradient tube (TGT) and filters. Irradiated fuel pellets were placed into a tungsten crucible, then loaded into the induction furnace. The temperature was raised continuously at a heating rate of 10 K/s to the targeted temperature and maintained for 500 s in a flowing argon gas atmosphere. The FPs and actinides released from the MOX fuels and deposited in the TGT and filters were quantified by gamma-ray spectrometry and inductively coupled plasma mass spectrometry (ICP-MS) analysis. Based on the analysis, the release rates of radionuclides from MOX fuels for FR were obtained and compared with literature data for light water reactor (LWR) fuels. The release rate coefficients of FPs obtained in this study were found to be similar to or lower than the literature values for LWR fuels. It was also found that the release rate coefficient data for actinides were within the range of variation of literature values for LWR fuels.
Aoyagi, Mitsuhiro; Uchibori, Akihiro; Kikuchi, Shin; Takata, Takashi; Ohno, Shuji; Ohshima, Hiroyuki
Nihon Kikai Gakkai Rombunshu (Internet), 86(883), p.19-00366_1 - 19-00366_8, 2020/03
Sodium fire is one of key issues in sodium-cooled fast reactor plant. JAEA has developed sodium fire analysis codes, such as AQUA-SF and SPHINCS, to evaluate the consequence of sodium fire events. This paper describes the PIRT (Phenomena Identification and Ranking Table) process for sodium fire events. Ranking table for important phenomena and an assessment matrix are completed. As a part of comprehensive validation based on the assessment matrix, experimental analyses using the AQUA-SF and SPHINCS codes for a sodium spray fire experiment Run-E1 show good agreement with the experimental result.
Aoyagi, Mitsuhiro; Uchibori, Akihiro; Kikuchi, Shin; Takata, Takashi; Ohno, Shuji; Ohshima, Hiroyuki
Nuclear Engineering and Design, 353, p.110240_1 - 110240_10, 2019/11
Times Cited Count：4 Percentile：50.24(Nuclear Science & Technology)JAEA has developed sodium fire analysis codes to evaluate the consequences of sodium fire events. This paper describes a PIRT (Phenomena Identification and Ranking Table) process for such events. Because a sodium fire event involves complex phenomena, the FOMs are specified through a factor analysis. Associated phenomena in a sodium fire event are identified through both element- and sequence-based phenomena analyses. The importance of each phenomenon is evaluated by considering the sequence-based analysis of associated phenomena related to the FOMs. Then, a ranking table is established. An assessment matrix of important phenomena and experiments is completed to confirm the sufficiency of experimental data for the validation of the models in the sodium fire analysis codes. Additional assessments are discussed specifically for the aerosol module and the CFD module in three-dimensional codes from a perspective of careful validation.
Aoyagi, Mitsuhiro; Takata, Takashi; Ohno, Shuji; Uno, Masayoshi*
Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 10 Pages, 2018/10
Heat radiation is one of dominant heat transfer process during a sodium fire event which is a concern in sodium-cooled fast reactor plants. This study aims to model radiation heat transfer from combusting droplets. Radiation energy transport on the combustion flame surface around a sodium droplet is formulated considering emission, absorption and scattering through a similar approach to the formulation of the wall boundary condition. The improved model is tested trough a simple verification analysis and a benchmark analysis on an upward sodium spray combustion experiment. As the result, overestimation of atmospheric temperature and pressure is mitigated by the improved model due to increase in heat transfer to structure.
Aoyagi, Mitsuhiro; Takata, Takashi; Ohno, Shuji; Uno, Masayoshi*
Nihon Kikai Gakkai Rombunshu (Internet), 84(859), p.17-00374_1 - 17-00374_13, 2018/03
no abstracts in English
Zhan, Y.*; Oya, Naoki*; Enoki, Koji*; Okawa, Tomio*; Ohno, Shuji; Aoyagi, Mitsuhiro; Takata, Takashi
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 6 Pages, 2017/07
It is important to set the amount of sodium droplet mechanistically for appropriate numerical evaluations of sodium leak and fire behavior in a sodium-cooled fast reactor plant. In the present work, fundamental experiments are performed to measure the splash ratio during the vertical water jet impact onto a horizontal wall. It is shown that the splash ratio can be correlated well as a function of the impact Weber number and dimensionless impact frequency.
Takata, Takashi; Ohno, Shuji; Tajima, Yuji*
Mechanical Engineering Journal (Internet), 4(3), p.16-00577_1 - 16-00577_11, 2017/06
Evaluation of accidental sodium leak, combustion, and its thermal consequence is one of the important issues to be assessed in the field of sodium-cooled fast reactor (SFR). The present paper deals with the sodium pool fire and subsequent heat transfer behavior in air atmosphere two-cell geometry both experimentally and analytically because such two-cell configuration is considered as a typical one to possess important characteristic of multi-compartment system seen in an actual plant. As a result of the numerical analysis using a lumped-parameter based zonal model safety analysis code SPHINCS, the applicability of the ventilation model implemented in SPHINCS has been demonstrated. It is also investigated that the buoyancy- driven ventilation is dominant in the experiment.
Aoyagi, Mitsuhiro; Uchibori, Akihiro; Kikuchi, Shin; Takata, Takashi; Ohno, Shuji; Ohshima, Hiroyuki
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Next Generation Nuclear Systems for Sustainable Development (FR-17) (USB Flash Drive), 10 Pages, 2017/06
The present PIRT process is aimed to identify key phenomena involved in sodium fire accidents that involve complex phenomena in sodium-cooled fast reactor plants. In this PIRT process, the figures of merit (FOMs) are specified through factor analysis. Associated phenomena are identified through the element- and sequence-based phenomena analyses. Importance of each associated phenomenon is evaluated by considering the sequence-based analysis of associated phenomena correlated with the FOMs. Then, we complete the ranking table through the factor and phenomenon analyses. An assessment matrix of important phenomena and experiments is completed finally for model validation.
Yi, Z.*; Oya, Naoki*; Enoki, Koji*; Okawa, Tomio*; Ohno, Shuji; Aoyagi, Mitsuhiro; Takata, Takashi
Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 7 Pages, 2016/11
It is important to set the amount of sodium droplet mechanistically for appropriate numerical evaluations of sodium leak and fire behavior in a sodium-cooled fast reactor plant. In the present work, fundamental experiments were performed to measure the splash ratio during the vertical water jet impact onto a horizontal wall. It was shown that the splash ratio can be correlated well as a function of the impact Weber number, the Strouhal number and the Ohnesorge number of the droplets impinging the liquid film.
Aoyagi, Mitsuhiro; Uchibori, Akihiro; Kikuchi, Shin; Takata, Takashi; Ohno, Shuji; Ohshima, Hiroyuki
Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 10 Pages, 2016/11
The present PIRT process was aimed to identify key phenomena involved in sodium fire accidents that involve complex phenomena in sodium-cooled fast reactor plants. In this PIRT process, the figures of merit (FOMs) were specified through factor analysis. Associated phenomena were identified through the element- and sequence-based phenomena analyses. Importance of each associated phenomenon was evaluated by considering the sequence-based analysis of associated phenomena correlated with the FOMs. Then, we complete the ranking table through the factor and phenomenon analyses.
Kojima, Saori*; Uchibori, Akihiro; Takata, Takashi; Ohno, Shuji; Fukuda, Takeshi*; Yamaguchi, Akira*
Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 8 Pages, 2016/11
Analytical evaluation on a self-wastage phenomenon at heat transfer tubes in the steam generator of sodium cooled fast reactors has been performed by using the sodium-water reaction analysis code SERAPHIM. In this study, a fluid-structure thermal coupling model was developed and incorporated in the SERAPHIM code to evaluate heat transfer between the sodium-side reacting flow and the outer surface of the heat transfer tube. The effect of the fluid-structure thermal coupling model on the temperature field was demonstrated through the numerical analyses.
Chikazawa, Yoshitaka; Kato, Atsushi*; Yamamoto, Tomohiko; Kubo, Shigenobu; Ohno, Shuji; Iwasaki, Mikinori*; Hara, Hiroyuki*; Shimakawa, Yoshio*; Sakaba, Hiroshi*
Nuclear Technology, 196(1), p.61 - 73, 2016/10
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)JSFR adopts double boundary for all sodium components. However, design measures are investigated for the secondary sodium fire inside the reactor building, which might be assumed as design extension conditions (DECs). Candidates of sodium fire measures in the secondary sodium systems such as sodium drain, nitrogen injection, pressure release valve, catch pan and leak sodium drain system have been compared from the view point of safety. Wide range of sodium fires in the steam generator room and air cooler have been analyzed evaluating performances of the candidate sodium fire measures.
Uchibori, Akihiro; Watanabe, Akira*; Takata, Takashi; Ohno, Shuji; Ohshima, Hiroyuki
Proceedings of OECD/NEA & IAEA Workshop on Application of CFD/CMFD Codes to Nuclear Reactor Safety and Design and their Experimental Validation (CFD4NRS-6) (Internet), 11 Pages, 2016/09
For assessment of the wastage environment under tube failure accident in a steam generator of sodium-cooled fast reactors, a mechanistic computer code called SERAPHIM calculating compressible multicomponent multiphase flow with sodium-water chemical reaction has been developed. The original SERAPHIM code is based on the finite difference method. In this study, unstructured mesh-based numerical method was developed and introduced into the SERAPHIM code to advance a numerical accuracy for the complex-shaped domain including multiple heat transfer tubes. Validity of the unstructured mesh-based SERAPHIM code was investigated through the analysis of an under expanded jet experiment. The calculated pressure profile showed good agreement with the experimental data. Numerical analysis of water vapor discharging into liquid sodium was also performed. It was demonstrated that the proposed numerical method could be applicable to evaluation of the sodium-water reaction phenomenon.
Ohno, Shuji; Takata, Takashi; Tajima, Yuji*
Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), 7 Pages, 2016/06
From various kinds of sodium fire situations postulated in SFR plants, the present paper treats the sodium pool fire and subsequent heat transfer behavior in an important air atmosphere two-cell geometry as one of the important cell configuration conditions. The detailed analysis and investigation of sodium fire and thermal-hydraulics in horizontally arranged two cells with an opening between them are made both from experimental measurement and from numerical simulation with a multi-cell sodium fire analysis code SPHINCS.
Yi, Z.*; Oya, Naoki*; Enoki, Koji*; Okawa, Tomio*; Ohno, Shuji; Aoyagi, Mitsuhiro
Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), 7 Pages, 2016/06
It is important to set the amount of sodium droplet mechanistically for appropriate numerical evaluations of sodium leak and fire behavior in a sodium-cooled fast reactor plant. In the present work, fundamental experiments were performed to measure the splash ratio during the vertical water jet impact onto a horizontal wall. It was shown that the splash ratio can be correlated well as a function of the impact Weber number and the Strouhal number of the droplets impinging the liquid film.
Kikuchi, Shin; Koga, Nobuyoshi*; Seino, Hiroshi; Ohno, Shuji
Journal of Nuclear Science and Technology, 53(5), p.682 - 691, 2016/05
Times Cited Count：14 Percentile：82.69(Nuclear Science & Technology)In a sodium-cooled fast reactor (SFR), if considering hypothetical severe accidental condition such as the steel liner failure of structural concrete caused by intensive leakage of liquid sodium (Na) coolant, the liquid sodium-concrete reaction (SCR) may take place. The major consequences of SCR are hydrogen release, energy release and concrete ablation. Thus, it is important to understand the phenomenology of SCR. As a part of a series of studies on SCR, this study focused on the reaction between sodium oxide (Na
O) and silica (SiO). Through thermoanalytical and X-ray diffraction measurements, it was revealed that NaO-SiO reaction to form sodium orthosilicate (Na
SiO) occurs at significantly lower temperature in comparison with Na-SiO reaction.
Tanaka, Masaaki; Ohno, Shuji; Ohshima, Hiroyuki
Nuclear Engineering and Design, 299, p.174 - 183, 2016/04
Times Cited Count：4 Percentile：39.97(Nuclear Science & Technology)A procedure called as V2UP (Verification and Validation plus Uncertainty quantification and Prediction) was made by referring to the existing guidelines on V&V and the methodologies of the safety assessment (CSAU, ISTIR, EMDAP). The V2UP consisted of five components as follows: (1) phenomena analysis with the Phenomena Identification and Ranking Table (PIRT) method, (2) implementation of the V&V, (3) design and rearrangement of experiments for the V&V, (4) uncertainty quantification in each problem and integration of uncertainties and (5) numerical prediction (estimation) for the target issue. Although the complete application of the procedure has not been performed at this moment, a flow chart of the V2UP procedure was described in this paper with recent results of the examinations.
Ohno, Shuji
Nihon Kikai Gakkai 2015-Nendo Nenji Taikai Koen Rombunshu (DVD-ROM), 4 Pages, 2015/09
A liquid metal sodium leak and pool fire experiment has provided the transient temperature data measured in more than 100 positions for atmospheric gas and structures. The experiment was performed in an air-filled 2-cell geometry that was made by partitioning the room of rectangular cell with an opening between the cells. The analyses of the measured data clarified the basic characteristics of sodium pool combustion and consequential heat and mass transfer in the cells. The data also suggested several features of multi-dimensional thermal-hydraulic behaviors that occurred in the experimental geometry during the fire. The measured data are expected to be utilized for the validation study of numerical sodium fire analysis tools.
Kobayashi, Jun; Tanaka, Masaaki; Ohno, Shuji; Ohshima, Hiroyuki; Kamide, Hideki
Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.6664 - 6677, 2015/08
Numerical simulation is recognized an essential tool for the physical phenomena analysis and plant design study of a sodium-cooled fast reactor (SFR). In order to enhance credibility of the numerical results in the activities for plant design by using numerical simulations, it is recognized that verification and validation (V&V) process is very important. In this study, experiments for planar triple parallel jets mixing phenomena conducted in JAEA were proposed as benchmark problems for the code validation in the area of thermal striping study in the SFR development.
Kikuchi, Shin; Koga, Nobuyoshi*; Seino, Hiroshi; Ohno, Shuji
Journal of Thermal Analysis and Calorimetry, 121(1), p.45 - 55, 2015/07
Times Cited Count：12 Percentile：44.91(Thermodynamics)In this study, the kinetic behavior of the sodium (Na)-silica (SiO) reaction was investigated for an assessment method of reactivity/stability of siliceous concrete against the sodium-concrete reaction (SCR) by postulating a severe accidental condition in the sodium-cooled fast reactor (SFR). The reaction behavior was tracked using a differential scanning calorimetry (DSC) equipped with a videoscope for viewing the changes in the sample during the reaction. From detail kinetic analysis, it was revealed that the kinetic results determined from the kinetic data at the maximum reaction rate can be interpreted as is for the major reaction stage. In addition, the k value at a constant temperature calculated using the Arrhenius parameters determined by the simplified Kissinger method can be used for the reactivity/stability assessment of the siliceous concrete in view of the kinetics of the major reaction stage of the Na-SiO reaction.
