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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.
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.
particle impinging on Al at 230 MeV/uFuruta, Toshimasa*; Uozumi, Yusuke*; Yamaguchi, Yuji; Iwamoto, Yosuke; Koba, Yusuke*; Velicheva, E.*; Kalinnikov, V.*; Tsamalaidze, Z.*; Evtoukhovitch, P.*
Journal of Nuclear Science and Technology, 61(2), p.230 - 236, 2024/02
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Charged particle production from particle fragmentation reactions was investigated experimentally by measurement of 230-MeV/u particles bombarding an aluminum target. Double differential cross sections were measured for each ejectile of p, d, t, 
He, and 
He at laboratory angles between 15 and 60 deg. The results of analyzed data found the following common characteristics: (1) spectra of proton- and neutron-emission are similar in high energy region at forward angle, (2) triton-to-He ratio of -breakup yield is 1:2, which is similar to lower incident energy experiment, and (3) the shape of broad peak formed by He and particles could be explained by the process with collision between induced particle and target nucleus.
Zablackaite, G.; Shiotsu, Hiroyuki; Kido, Kentaro; Sugiyama, Tomoyuki
Nuclear Engineering and Technology, 56(2), p.536 - 545, 2024/02
Times Cited Count：0Mohamad, A. B.; Udagawa, Yutaka
Nuclear Technology, 210(2), p.245 - 260, 2024/02
Times Cited Count：1 Percentile：72.91(Nuclear Science & Technology)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.
Riyana, E. S.; Okumura, Keisuke; Sakamoto, Masahiro; Matsumura, Taichi; Terashima, Kenichi; Kanno, Ikuo
Journal of Nuclear Science and Technology, 8 Pages, 2024/00
Times Cited Count：0 Percentile：0.18(Nuclear Science & Technology)
Re in the keV energy regionKatabuchi, Tatsuya*; Sato, Yaoki*; Takebe, Karin*; Igashira, Masayuki*; Umezawa, Seigo*; Fujioka, Ryo*; Saito, Tatsuhiro*; Iwamoto, Nobuyuki
Journal of Nuclear Science and Technology, 6 Pages, 2024/00
Times Cited Count：0 Percentile：0.18(Nuclear Science & Technology)Tamura, Koji; Nakanishi, Ryuzo; Oba, Hironori; Karino, Takahiro; Shibata, Takuya; Taira, Takunori*; Wakaida, Ikuo
Journal of Nuclear Science and Technology, 8 Pages, 2024/00
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Li, F.; Mihara, Takeshi; Udagawa, Yutaka
Journal of Nuclear Science and Technology, 11 Pages, 2024/00
Times Cited Count：0Soma, Shu; Ishigaki, Masahiro*; Abe, Satoshi; Shibamoto, Yasuteru
Nuclear Engineering and Technology, 10 Pages, 2024/00
Seki, Akiyuki; Yoshikawa, Masanori; Nishinomiya, Ryota*; Okita, Shoichiro; Takaya, Shigeru; Yan, X.
Nuclear Technology, 12 Pages, 2024/00
Times Cited Count：0 Percentile：0.18(Nuclear Science & Technology)Two types of deep neural network (DNN) systems have been constructed with the intent to assist safety operation of a nuclear power plant. One is a surrogate system (SS) that can estimate physical quantities of a nuclear power plant in a computational time of several orders less than a physical simulation model. The other is an abnormal situation identification system (ASIS) that can estimate the state of the disturbance causing an anomaly from physical quantities of a nuclear power plant. Both systems are trained and tested using data obtained from the analytical code for incore and plant dynamics (ACCORD), which reproduces the steady and dynamic behavior of the actual high Temperature engineering test reactor (HTTR) under various scenarios. The DNN models are built by adjusting, the main hyperparameters. Through these procedures, these systems are shown able to perform with a high degree of accuracy.
McGrady, J.; Kumagai, Yuta; Watanabe, Masayuki; Kirishima, Akira*; Akiyama, Daisuke*; Kimuro, Shingo; Ishidera, Takamitsu
Journal of Nuclear Science and Technology, 60(12), p.1586 - 1594, 2023/12
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Nakayama, Shinsuke
Journal of Nuclear Science and Technology, 60(12), p.1447 - 1453, 2023/12
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)The d+Be neutron source is a candidate for transportable neutron source for on-site nondestructive inspection of infrastructure facilities such as bridges, tunnels and so on. The applicability of the d+Be neutron source to a transportable fast neutron source is explored by Monte Carlo particle transport simulations with PHITS and JENDL-5. The simulation results show that by increasing the shielding thickness by about 1.5 times, it is possible to realize the d+Be neutron source with the comparable performance to another candidate, the 2.5-MeV p+Li neutron source, at lower beam energy.
Fukuda, Kodai; Yamane, Yuichi
Journal of Nuclear Science and Technology, 60(12), p.1514 - 1525, 2023/12
Times Cited Count：1 Percentile：72.91(Nuclear Science & Technology)This study aims to clarify the effect of fuel particle radius on the criticality transient behavior and the total number of fissions in water-moderated solid fuel dispersion systems. Neutronics/thermal hydraulics-coupled kinetics analysis was performed in a hypothetical fuel debris system, where small fuel particles aggregate in water and become supercritical. Results showed that the number of fissions is 10 times larger when the fuel particle radius is reduced by one order of magnitude under conditions where heat transfer, i.e. from fuel to water, is emphasized. Moreover, there is a possibility that lower reactivity could give a larger number of fissions when the fuel particle size is very small. In addition, the number of fissions may be overestimated or underestimated to an unexpected extent unless appropriate fuel particle size is set on the analysis.
Tagami, Hirotaka; Ishida, Shinya; Tobita, Yoshiharu
Journal of Nuclear Science and Technology, 60(12), p.1548 - 1562, 2023/12
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)In a design of future Sodium-cooled Fast Reactor, there is a demand for evaluation of sequences and consequences of core disruptive accidents. Future SFRs include a unique core design with axially or horizontally heterogeneous core arrangement having complex fuel isotope distribution. A new model to flexibly represent fuel isotope distribution, called the Pu-vector model, has been developed in this study for inclusion in the SIMMER-III and SIMMER-IV codes (simply called as SIMMER). The model calculates movement of individual fuel isotopes, assuming they always accompany the convecting fuel in the fluid-dynamics model. The accuracy of the Pu-vector model was confirmed by comparing with the standard Monte Carlo static neutronics calculation. The new model can improve some of the limitations in the current SIMMER code, in which the fuel isotopes are represented only by two groups, fertile and fissile fuels. Assignment of a number of fuel isotopes to the two groups requires a detailed examination of different combinations of fuel isotopes to determine an optimized combination. The Pu-vector model can eliminate this complicated procedure to be performed prior to a SIMMER analysis, and more importantly provides accurate spatial distribution of fuel isotopes and thus will improve the applicability of SIMMER to the analyses of future large heterogeneous reactors.
Takasaki, Koji; Yasumune, Takashi; Yamaguchi, Yukako; Hashimoto, Makoto; Maeda, Koji; Kato, Masato
Journal of Nuclear Science and Technology, 60(11), p.1437 - 1446, 2023/11
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)The aerodynamic radioactive median diameter (AMAD) is necessary information to assess the internal exposure. On June 6, 2017, at a plutonium handling facility in Oarai site of Japan Atomic Energy Agency (JAEA), during the inspection work of a storage container that contains nuclear fuel materials, accidental contamination occurred and five workers inhaled radioactive materials including plutonium. Some smear papers and an air sampling filter were measured with the imaging plate, and we conservatively estimated minimum AMADs for two cases, plutonium nitrate and plutonium dioxide. As a result of AMAD estimation, even excluding a giant particle of a smear sample, the minimum AMADs of plutonium nitrate from smear papers were 4.3 - 11.3 
m and those of plutonium dioxide were 5.6 - 14.1 m. Also, the minimum AMAD of plutonium nitrate from an air sampling filter was 3.0 m and that of plutonium dioxide was 3.9 m.
Tsuchiya, Harufumi; Toh, Yosuke; Ozu, Akira; Furutaka, Kazuyoshi; Kitatani, Fumito; Maeda, Makoto; Komeda, Masao
Journal of Nuclear Science and Technology, 60(11), p.1301 - 1312, 2023/11
Times Cited Count：1 Percentile：72.91(Nuclear Science & Technology)
unheated rod bundleNagatake, Taku; Yoshida, Hiroyuki
Journal of Nuclear Science and Technology, 60(11), p.1417 - 1430, 2023/11
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)In recent years, computational fluid dynamics (CFD) codes have been used to evaluate two-phase flow behavior inside a fuel bundle for nuclear core design and accident management. A space-time distribution of void fraction and interfacial velocity in bundle systems at high temperatures and pressures, are important for validation of two-phase flow CFD codes. However, it is difficult to obtain a space-time distribution of void fraction and interfacial velocity in a bundle system at high temperature and pressure conditions. We have so far developed an experimental apparatus with a unheated rod bundle by adapting a through-rod WMS to measure distributions of a void fraction and an interfacial velocity in high pressure conditions. We newly measured distributions of the void fraction and interfacial velocity in the water-vapor system under high pressure up to 2.6 MPa by the developed apparatus. It has been confirmed that reasonable results were obtained by the experimental apparatus.
