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Yoshikawa, Ryuji; Imai, Yasutomo*; Kikuchi, Norihiro; Tanaka, Masaaki; Ohshima, Hiroyuki
Nuclear Technology, 210(5), p.814 - 835, 2024/05
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)In the study of safety enhancement on advanced sodium-cooled fast reactor, it is essential to clarify the thermal-hydraulics under various operation conditions in a fuel assembly (FA) with the wire-wrapped fuel pins to assess the structural integrity of the fuel pin. A finite element thermal-hydraulics analysis code named SPIRAL has been developed to analyze the detailed thermal-hydraulics phenomena in a FA. In this study, the numerical simulations of the 37-pin bundle sodium experiments at different Re number conditions, including a transitional condition between laminar and turbulent flows and turbulent flow conditions, were performed to validate the hybrid turbulence model equipped in SPIRAL. The temperature distributions predicted by SPIRAL was consistent with those measured in the experiments. Through the validation study, the applicability of the hybrid turbulence model in SPIRAL to thermal-hydraulic evaluation of sodium-cooled FA in the wide range of Re number was confirmed.
Uwaba, Tomoyuki; Ito, Masahiro*; Ishitani, Ikuo*
JAEA-Technology 2023-006, 36 Pages, 2023/05
JAEA-Technology-2023-006.pdf:3.45MB
The BAMBOO code developed by the Japan Atomic Energy Agency is a computer code to analyze fuel pin bundle deformation in a fast reactor wire-spaced type fuel pin bundle subassembly. In this study we developed an analysis model to consider friction at the contact points between adjacent fuel pins, and at these between outermost fuel pins and a duct that are due to bundle-duct interaction. This model deals with friction forces at contact points in the contact and separation analysis of the code, and employs a convergent calculation where contact forces are gradually determined to avoid numerical instability when the friction occurs. Analyses of BAMBOO with the model showed very slight effects on the onset of contact between outer most pins and a duct, and on directions of pin displacements, within the range of practical friction coefficients.
Ono, Ayako; Yamashita, Susumu; Suzuki, Takayuki*; Yoshida, Hiroyuki
Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 16 Pages, 2022/03
JAEA is developing the methodology to predict the critical heat flux based on a mechanism in order to reduce the cost for full mock-up test. The evaluation method based on a mechanism is expected to be able to predict in the wide range of parameter under the unexpected conditions including the severe accident. In this study, the JUPITER code developed by JAEA is examined to apply for the two-phase flow simulation of LWR fuel assembly with the spacer grid. The benchmark data of single-phase flow in the bundle with the spacers by KAERI were used to validate the simulation result by JUPITER. Moreover, the single-phase flow simulation was conducted by another simulation method, STAR-CCM+, as a supplemental analysis to consider the effect of the different simulation methods. Finally, the two-phase flow simulation for the bundle with the spacer was conducted by JUPITER. The effect of the spacer with a vane on the bubble behavior is discussed.
Uwaba, Tomoyuki; Nemoto, Junichi*; Ito, Masahiro*; Ishitani, Ikuo*; Doda, Norihiro; Tanaka, Masaaki; Otsuka, Satoshi
Nuclear Technology, 207(8), p.1280 - 1289, 2021/08
Times Cited Count：3 Percentile：34.82(Nuclear Science & Technology)Computer codes for irradiation behavior analysis of a fuel pin and a fuel pin bundle and for coolant thermal hydraulics analysis were coupled into an integrated code system. In the system, each code provides data required by other codes and the analyzed results are shared among them. The system allows for the synthesizing of analyses of thermal, chemical and mechanical behaviors in a fuel subassembly under irradiation. A test analysis was made for a fuel subassembly containing a mixed oxide fuel pin bundle irradiated in a fast reactor. The results of the analysis were presented with transverse cross-sectional images of the fuel subassembly and three-dimensional images of a fuel pin and fuel pin bundle models. For detailed evaluation, various irradiation behaviors of all fuel pins in the subassembly were analyzed and correlated with irradiation conditions.
4 simulated bundleOno, Ayako; Yamashita, Susumu; Suzuki, Takayuki*; Yoshida, Hiroyuki
Mechanical Engineering Journal (Internet), 7(3), p.19-00583_1 - 19-00583_12, 2020/06
JAEA is implementing the 3D detailed nuclear-thermal-coupled analysis code to analyze the transition state of the core and to reduce the likelihood of the design. In the development plan, the computational fluid dynamics code based on the VOF method, JUPITER, is applied for TH part of the 3D detailed nuclear-thermal-coupled analysis code.
Uwaba, Tomoyuki; Ohshima, Hiroyuki; Ito, Masahiro*
Nuclear Engineering and Design, 317, p.133 - 145, 2017/06
Times Cited Count：9 Percentile：65.3(Nuclear Science & Technology)The coupled numerical analysis of mechanical and thermal behaviors was performed for a wire-wrap fuel pin bundle subassembly irradiated in a fast reactor. For the analysis, the fuel pin bundle deformation analysis code BAMBOO and the thermal hydraulics analysis code ASFRE exchanged the deformation and temperature analysis results through the iterative calculations to attain convergence corresponding to the static balance between deformation and temperature. The analysis by the coupled code system showed that radial distribution of coolant temperatures in a subassembly tended to be flattened as a result of the fuel pin bundle deformation governed by cladding void swelling and irradiation creep. Such temperature distribution was slightly analyzed as a result of the small bowing of the fuel pins due to the cladding-wire interaction even when no bundle-duct interaction occurred. The effect of the spacer wire-pitch on deformation and thermal hydraulics was also investigated in this study.
Higashiuchi, Atsushi; Ishimi, Akihiro; Katsuyama, Kozo; Uwaba, Tomoyuki; Ichikawa, Shoichi
JAEA-Technology 2015-057, 72 Pages, 2016/03
JAEA-Technology-2015-057.pdf:36.91MB
Bundle-duct interaction (BDI) in fast reactors (FRs) is one of the limiting factors for burnup. To study the high performance fuel for FR fuel, it is important to establish the method to predict accurately the BDI behavior for the fuel assembly of large-diameter fuel pins. Therefore, it was adopted a new method that the bundle compression test apparatus is placed outside the cell, the bundle specimen is put in the airtight container for contamination prevention, and the bundle specimen is carried in the cell for internal observation by X-ray CT examination apparatus. From the result of this test, it was confirmed that the new method of out-of-pile bundle compression test is carried out as it was before. The results of this test are available to study integrity assessment of fast reactor fuel, validation of the BDI analysis code and substantiation of the safety design guidelines of fast reactor. In addition, it is possible to reflect in the BDI behavior evaluation for "ASTRID".
Yoshida, Hiroyuki; Tamai, Hidesada; Onuki, Akira; Takase, Kazuyuki; Akimoto, Hajime
Nuclear Engineering and Technology, 38(2), p.119 - 128, 2006/04
The reduced-moderation water reactor core adopts a hexagonal tight-lattice arrangement. In the core, there is no sufficient information about the effects of the gap spacing and grid spacer configuration on the flow characteristics. Thus, we start to develop a predictable technology for thermal-hydraulic performance of the core using an advanced numerical simulation technology. As a part of this technology development, we are developing a two-phase flow simulation code TPFIT with an advanced interface tracking method. The vector and parallelization of the code was conducted to fit the large-scale simulations. The numerical results applied to large-scale water-vapor two-phase flow in tight lattice rod bundles are shown and compared with experimental results. In the results, a tendency of the predicted void fraction distribution in horizontal plane agreed with the measured values including the bridge formation of the liquid at the position of adjacent fuel rods where an interval is the narrowest.
Takase, Kazuyuki; Yoshida, Hiroyuki; Ose, Yasuo*; Akimoto, Hajime
Computational Fluid Dynamics 2004, p.649 - 654, 2006/00
no abstracts in English
Takase, Kazuyuki; Yoshida, Hiroyuki; Ose, Yasuo*; Tamai, Hidesada
Proceedings of 6th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operations and Safety (NUTHOS-6) (CD-ROM), 14 Pages, 2004/10
no abstracts in English
W.Hering*; Minato, Kazuo; Nagase, Fumihisa
Nuclear Technology, 102, p.100 - 115, 1993/04
Times Cited Count：2 Percentile：29.78(Nuclear Science & Technology)no abstracts in English
Kumamaru, Hiroshige; Kukita, Yutaka
Nucl. Eng. Des., 144, p.257 - 268, 1993/00
Times Cited Count：1 Percentile：18.76(Nuclear Science & Technology)no abstracts in English
Kumamaru, Hiroshige; Kukita, Yutaka
ANP 92: Proc. of the Int. Conf. on Design and Safety of Advanced Nuclear Power Plants,Vol. 3, p.24.4-1 - 24.4-7, 1992/00
no abstracts in English
; *
JAERI 1293, 82 Pages, 1984/11
no abstracts in English
; ; *
JAERI-M 9495, 79 Pages, 1981/05
no abstracts in English
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JAERI-M 9319, 85 Pages, 1981/02
no abstracts in English
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Nucl.Eng.Des., 65, p.63 - 69, 1981/00
Times Cited Count：5 Percentile：58.06(Nuclear Science & Technology)no abstracts in English
Doda, Norihiro; Ohshima, Hiroyuki; Kamide, Hideki; Watanabe, Osamu*
no journal, ,
We have developed a simplified model for Highest Cladding Temperature Evaluation in Wire-wrapped Fuel Rod Bundle under Natural Circulation Decay Heat Removal. We have demonstrated the case of 271-pin fuel assembly in the external power loss event with the simplified model, and the results were in good agreement with the results by using the subchannel analysis code. Some uncertainties in the fuel assembly, e.g. the variations of fuel pin power and fuel pin position, were considered by setting boundary conditions. The computational cost of the simplified model was about one-thousandth compared with the subchannel analysis model. The simplified model can be applied without unrealistic computational cost to the statistical evaluation of safety criterion under natural circulation decay heat removal, which requires a large number of transient analyses.
