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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.
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.
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.
