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Nakamichi, Shinya; Sunaoshi, Takeo*; Hirooka, Shun; Vauchy, R.; Murakami, Tatsutoshi
Journal of Nuclear Materials, 595, p.155072_1 - 155072_11, 2024/07
Frazer, D.*; Saleh, T. A.*; Matsumoto, Taku; Hirooka, Shun; Kato, Masato; McClellan, K.*; White, J. T.*
Nuclear Engineering and Design, 423, p.113136_1 - 113136_7, 2024/07
Nanoindentation based techniques can be employed on minute volumes of material to measure mechanical properties, including Young's modulus, hardness, and creep stress exponents. In this study, (U,Ce)O solid solutions samples are used to develop elevated temperature nanoindentation and nanoindentation creep testing methods for use on mixed oxide fuels. Nanoindentation testing was performed on 3 separate (Ux-1,Cex)O compounds ranging from x equals 0.1 to 0.3 at up to 800 C: their Young's modulus, hardness, and creep stress exponents were evaluated. The Young's modulus decreases in the expected linear manner while the hardness decreases in the expected exponential manner. The nanoindentation creep experiments at 800 C give stress exponent values, n=4.7-6.9, that suggests dislocation motion as the deformation mechanism.
Hong, Z.*; Ahmed, Z.*; Pellegrini, M.*; Yamano, Hidemasa; Erkan, N.*; Sharma, A. K.*; Okamoto, Koji*
Progress in Nuclear Energy, 171, p.105160_1 - 105160_13, 2024/06
In this study, it is found that the eutectic reaction between BC powder and stainless steel (SS) is considerably more rapid than that between the BC pellet and SS. The derived reaction rate constant values for powder and pellet cases are consistently based on the reference values. Also, a composition analysis using SEM/EDS was conducted for the detailed microstructures of the powder and pellet samples. In the powder case, only one thick layer is found as the reaction layer consisting of (Fe, Cr)B precipitate, including BC powder. In the pellet case, two layers are found in the reaction layer.
Kato, Masato; Oki, Takumi; Watanabe, Masashi; Hirooka, Shun; Vauchy, R.; Ozawa, Takayuki; Uwaba, Tomoyuki; Ikusawa, Yoshihisa; Nakamura, Hiroki; Machida, Masahiko
Journal of the American Ceramic Society, 107(5), p.2998 - 3011, 2024/05
Times Cited Count:0 Percentile:0.01(Materials Science, Ceramics)Yoshikawa, Ryuji; Imai, Yasutomo*; Kikuchi, Norihiro; Tanaka, Masaaki; Ohshima, Hiroyuki
Nuclear Technology, 210(5), p.814 - 835, 2024/05
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.
Namie, Masanari; Saito, Junichi
Computational Materials Science, 239, p.112963_1 - 112963_7, 2024/04
Imagawa, Yuya; Toyota, Kodai; Onizawa, Takashi; Kato, Shoichi
JAEA-Testing 2023-004, 76 Pages, 2024/03
This manual describes the methods for conducting material tests in air, argon gas, and sodium, and for organizing the data obtained, as a part of the development of high-temperature structural design technology for fast reactors. This manual reflects the revision of test methods in Japanese Industrial Standards (JIS) to the "FBR Metallic Materials Test Manual, PNC TN241 77-03" published in 1977 and the "FBR Metallic Materials Test Manual (Revised Edition), JNC TN9520 2001-001" published in 2001. Also, it was written with reference to the recommended room temperature / elevated temperature tensile test method by the Japan Society of Mechanical Engineers (JSME) and the test standard for the elevated-temperature low-cycle fatigue test method by the Society of Materials Science, Japan (JSMS), which are the standard for material test methods in the domestic academic society.
Endo, Tomohiro*; Maruyama, Shuhei; Yamamoto, Akio*
Journal of Nuclear Science and Technology, 61(3), p.363 - 374, 2024/03
Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)Uncertainty quantification (UQ) of the neutron multiplication factor is important to investigate the appropriate safety margin for a target system. Although the random sampling method is a practical and useful UQ method, a large computational cost is required to reduce the statistical error of the estimated uncertainty. Furthermore, if an input variable follows a normal distribution with a large standard deviation, the perturbed input variable by the random sampling method may become a physically inappropriate or negative value. To address these issues for the efficient and robust UQ, a modified deterministic sampling method using the simplex ensemble and the scaling method is proposed. The features of the proposed method are summarized as follows: The sample size is (r+2), where r corresponds to the effective rank of the covariance matrix between the input variables; depending on a situation of target UQ, the amounts of perturbations for the input parameters can be arbitrarily given by the scaling factor method; the scaling factor can be updated to avoid physically inappropriate in the perturbed input variables. The effectiveness of the proposed method is demonstrated through the UQ of the neutron multiplication factor due to fuel manufacturing uncertainties for a typical PWR pin-cell burnup calculation.
Futagami, Satoshi; Ando, Masanori; Yamano, Hidemasa
Transactions of the 27th International Conference on Structural Mechanics in Reactor Technology (SMiRT 27) (Internet), 10 Pages, 2024/03
Yamano, Hidemasa; Futagami, Satoshi; Ando, Masanori; Kurisaka, Kenichi
Transactions of the 27th International Conference on Structural Mechanics in Reactor Technology (SMiRT 27) (Internet), 11 Pages, 2024/03
In this study, the dynamic structural analysis of the reactor vessel for excessive earthquake using the FINAS/STAR code has shown the elephant foot buckling deformation and calculated the cumulative fatigue failure fraction. Using the calculation results, this paper describes the fragility curve using the safety factor method, indicating the significantly improved curve compared the previous one.
Yoshimura, Kazuo; Doda, Norihiro; Igawa, Kenichi*; Uwaba, Tomoyuki; Tanaka, Masaaki; Nemoto, Toshiyuki*
Transactions of the 27th International Conference on Structural Mechanics in Reactor Technology (SMiRT 27) (Internet), 8 Pages, 2024/03
To investigate possibility of the insertion of the reactivity by the deflection of the upper core support plate, structural mechanics analyses of the domain consisting of the fuel assemblies and core support plates and evaluation of the reactivity due to the inclination of the fuel assemblies in EBR-II were carried out. As a result, it was indicated that the upper core support plate deflected downward larger at the low flowrate condition than that at the high flowrate condition and positive reactivity was inserted due to the inclination of the fuel assemblies at the low flowrate condition.
Kasahara, Naoto*; Yamano, Hidemasa; Nakamura, Izumi*; Demachi, Kazuyuki*; Sato, Takuya*; Ichimiya, Masakazu*
Transactions of the 27th International Conference on Structural Mechanics in Reactor Technology (SMiRT 27) (Internet), 8 Pages, 2024/03
In this study, we propose failure mitigation methods by application of passive safety structures. The idea of the passive safety structures was applied to next generation fast reactors under high temperature conditions and excessive earthquake conditions.
Sun, G.*; Zhan, Y.*; Okawa, Tomio*; Aoyagi, Mitsuhiro; Uchibori, Akihiro; Okano, Yasushi
Experimental Thermal and Fluid Science, 151, p.111095_1 - 111095_15, 2024/02
Times Cited Count:1 Percentile:0.01(Thermodynamics)Hamase, Erina; Ohgama, Kazuya; Kawamura, Takumi*; Doda, Norihiro; Tanaka, Masaaki; Yamano, Hidemasa
Annals of Nuclear Energy, 195, p.110157_1 - 110157_14, 2024/01
Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)To validate the fast reactor plant dynamics analysis code Super-COPD for the loss of flow without scram (LOFWOS) event, we participated in the IAEA benchmark for the LOFWOS test No.13 performed at the FFTF as one of the passive safety demonstration test. In the blind phase, there were challenges to reproduce outlet temperatures of fuel assemblies and the total reactivity. To improve the evaluation accuracy of them, the whole core model considering the radial heat transfer and interwrapper flow and the simplified assembly bowing reactivity model were introduced. As a result of the final phase, the second peak of outlet temperatures was reproduced successfully, and the total reactivity could generally follow the measured data. Super-COPD was validated for the LOFWOS event.
Mikami, Nao*; Ueki, Yoshitaka*; Shibahara, Masahiko*; Aizawa, Kosuke; Ara, Kuniaki*
International Journal of Multiphase Flow, 171, p.104688_1 - 104688_13, 2024/01
Times Cited Count:0 Percentile:0(Mechanics)Sun, Y.*; Takatani, Tomoya*; Muta, Hiroaki*; Fujieda, Shun*; Kondo, Toshiki; Kikuchi, Shin; Kargl, F.*; Oishi, Yuji*
International Journal of Thermophysics, 45(1), p.11_1 - 11_19, 2024/01
Times Cited Count:0 Percentile:0.07(Thermodynamics)no abstracts in English
Horii, Yuta; Hirooka, Shun; Uno, Hiroki*; Ogasawara, Masahiro*; Tamura, Tetsuya*; Yamada, Tadahisa*; Furusawa, Naoya*; Murakami, Tatsutoshi; Kato, Masato
Journal of Nuclear Materials, 588, p.154799_1 - 154799_20, 2024/01
Times Cited Count:1 Percentile:72.91(Materials Science, Multidisciplinary)The thermal conductivities of near-stoichiometric (U,Pu,Am)O doped with NdO/SmO, which is major fission product (FP) generated by a uranium-plutonium mixed oxides (MOX) fuel irradiation, as simulated fission products are evaluated at 1073-1673 K. The thermal conductivities are calculated from the thermal diffusivities that are measured using the laser flash method. To evaluate the thermal conductivity from a homogeneity viewpoint of Nd/Sm cations in MOX, the specimens with different homogeneity of Nd/Sm are prepared using two kinds of powder made by ball-mill and fusion methods. A homogeneous Nd/Sm distribution decreases the thermal conductivity of MOX with increasing Nd/Sm content, whereas heterogeneous Nd/Sm has no influence. The effect of Nd/Sm on the thermal conductivity is studied using the classical phonon transport model (A+BT). The dependences of the coefficients A and B on the Nd/Sm content (C and C, respectively) are evaluated as: A(mK/W)=1.70 10 + 0.93C + 1.20C, B(m/W)=2.39 10.
Vauchy, R.; Matsumoto, Taku; Hirooka, Shun; Uno, Hiroki*; Tamura, Tetsuya*; Arima, Tatsumi*; Inagaki, Yaohiro*; Idemitsu, Kazuya*; Nakamura, Hiroki; Machida, Masahiko; et al.
Journal of Nuclear Materials, 588, p.154786_1 - 154786_13, 2024/01
Times Cited Count:1 Percentile:72.91(Materials Science, Multidisciplinary)Maruyama, Shuhei; Endo, Tomohiro*; Yamamoto, Akio*
Journal of Nuclear Science and Technology, 61(1), p.31 - 43, 2024/01
Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)This study investigated the feasibility of reducing the uncertainty associated with fast-reactor-core design by sharing an experimental database between different fields (e.g., reactor physics and radiation shielding) using data assimilation techniques. As the first step in this study, we focused on the ORNL sodium shielding experiment and investigated the possibility of using the experimental data to reduce the uncertainty in sodium void reactivity (SVR), which is the most important safety parameter for sodium-cooled fast reactors. A sensitivity analysis based on the Generalized Perturbation Theory was performed for the sodium shielding experiment. Using the sensitivity coefficients evaluated here and those of the sodium void reactivity previously evaluated by the JAEA, we showed that sodium shielding experimental data can contribute to the uncertainty reduction of SVR by adopting the cross-section adjustment method. Based on this study, the uncertainty reduction effect is expected to be significant, especially for SVR dominated by neutron-leakage phenomena. Although new reactor physics experimental data on SVR may be difficult to obtain, the results of this study suggest that data from sodium shielding experiments can partially substitute for this role. This study demonstrated the value of the mutual use of integral experimental data in fast reactor designs.
Kawaguchi, Munemichi; Hirakawa, Yasushi; Sugita, Yusuke; Yamaguchi, Yutaka
Nuclear Technology, 210(1), p.55 - 71, 2024/01
Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)This study has developed an estimation method for residual sodium film and sodium lumps on dummy fuel pins in Monju and demonstrated sodium draining behavior through gaps among the pins, experimentally. The amounts of the residual sodium on the surface of the pins were measured using the three-type test specimens: (a) single pin, (b) 7-pin assembly, and (c) 169-pin assembly. The experiments revealed that the withdrawal speed of the pins and improvement of the sodium wetting increased drastically the amounts of the residual sodium. Furthermore, the experiments using the 169-pin assembly measured the practical amounts of the residual sodium in the dummy fuel assembly of short length and demonstrated sodium draining behavior through the dummy fuel assembly. The estimation method includes four models: a viscosity flow model, Landau-Levich-Derjaguin (LLD) model, an empirical equation related to the Bretherton model, and a capillary force model in a tube. The calculation predicted comparable amounts of the residual sodium with the experiments. An uncertain of the sodium wetting effects were close to 1.8 times the estimation values of the LLD model. With this estimation method, the amounts of the residual sodium on the unloaded Monju dummy fuel assembly can be evaluated.