CFD analysis of the loss coefficient for a 90° bend in rolling motion

The local loss coefficient for a 90° bend in rolling motion is investigated with CFD code FLUENT. The calculation results are validated with experimental and theoretical results in steady state. The effect of spanwise and transverse additional forces on the bend loss is significant. The effects of additional forces on the bend loss are mainly embodied in the downstream section. The oscillation of bend loss caused by the spanwise and transverse additional forces is very regular while that caused by velocity oscillation is very irregular. The effect of velocity oscillation on the bend loss is significant in rolling motion with low Reynolds number. But the variation of bend loss coefficient with velocity oscillating period is very limited.     

Nuclear alternating parity bands and transition rates in a model of coherent quadrupole–octupole motion in neutron-rich barium isotopes

Deformed even–even nuclei Barium isotopes with quadrupole–octupole deformations are investigated on the basis of a collective model.The model describes energy levels of the yrast band with alternating parity in the neutron-rich~(140;142;144;146;148)Ba.The structure of the alternating parity bands is examined by odd–even (△I=1) staggering diagrams.An analytical method of the collective model is proposed for the calculation of E2 transition probabilities in alternating spectra of the nuclei~(140;142;144;146)Ba.     

Neutronics analysis of the power flattening and minor actinides burning in a thorium-based fusion–fission hybrid reactor blanket

A neutronics analysis has been performed for a thorium fusion breeder with a special task of burning minor actinide ²³⁷Np, ²⁴¹Am, ²⁴³Am, and ²⁴⁴Cm, and production of ²³³U for the future PWR application. Under a first wall fusion neutron wall loading of 0.1 MW/m² by a plant factor of 100%, preliminary neutronics calculations have been performed using the one-dimensional transport and burnup calculation code BISONC and the Monte Carlo transport code MCNP. To obtain a quasi-constant nuclear heat production density, 11 fuel rods containing the mixture of ThO₂ and minor actinides are placed in a radial direction in the fissile zone where ThO₂ is mixed with variable amounts of minor actinides. Calculation results show that the tritium breeding ratio is greater than 1.05 for both investigated Cases A and B, and the hybrid reactor is self-sufficient in the tritium required for the (DT) fusion driver in those models during the operation period. The blanket energy multiplication factor M, varies between 13.8 and 29.6 depending on the fuel types at the end of the operation period. The peak-to-average fission power density ratio (Γ) is less than 1.66 and 1.68 for both Cases A and B, respectively during the operation time. After 720 days of plant operation, the accumulated ²³³U is 1277 and 1725 kg in the blanket for the Cases A and B, respectively.     

Creep damage characterization of UNS N10003 alloy based on a numerical simulation using the Norton creep law and Kachanov–Rabotnov creep damage model

The calculation of inelastic creep damage is important for the structural integrity evaluation of the elevated temperature structure in a thorium molten salt reactor(TMSR). However, a creep damage theory model and numerical simulation method have not been proposed for the key materials(UNS N10003 alloy) in the TMSR. In this study, creep damage characterization of UNS N10003 alloy is investigated using the Norton creep law and Kachanov–Rabotnov(K–R) creep damage model. First, the creep experimental data of the UNS N10003 alloy at 650 °C were adopted to fit the material constants of the two models. Then, the creep damage behavior of the UNS N10003 alloy was analyzed and discussed under uniaxial and multi-axial stress states. The results indicated that the K–R creep damage model is more suitable for the UNS N10003 alloy than the Norton model. Finally, the numerical simulation method was developed by a user-defined UMAT subroutine and subsequently verified through a finite element analysis(FEA). The FEA results were in agreement with the theoretical solutions. This study provides an effective method for the inelastic creep damage analysis of the elevated temperature structure in the TMSR.     

Early construction and operation of the highly contaminated water treatment system in Fukushima Daiichi Nuclear Power Station (III) – a unique simulation code to evaluate time-dependent Cs adsorption/desorption behavior in column system

A simulation code was developed to evaluate the performance of the cesium adsorption instrument operating in Fukushima Daiichi Nuclear Power Station. Since contaminated water contains seawater whose salinity is not constant, a new model was introduced to the conventional zeolite column simulation code to deal with the variable salinity of the seawater. Another feature of the cesium adsorption instrument is that it consists of several columns arranged in both series and parallel. The spent columns are replaced in a unique manner using a merry-go-round system. The code is designed by taking those factors into account. Consequently, it enables the evaluation of the performance characteristics of the cesium adsorption instrument, such as the time history of the decontamination factor, the cesium adsorption amount in each column, and the axial distribution of the adsorbed cesium in the spent columns. The simulation is conducted for different operation patterns and its results are given to Tokyo Electric Power Company (TEPCO) to support the optimization of the operation schedule. The code is also used to investigate the cause of some events that actually occurred in the operation of the cesium adsorption instrument.     

Measurements of activation cross sections of (n,p) and (n, α) reactions in the energy range of 3.5–5.9 MeV using a deuterium gas target

Activation cross sections of (n, p) and (n, α) reactions were measured by means of the activation method in the neutron energy range of 3.5–5.9 MeV using a deuterium gas target. The irradiated target isotopes were ²⁷Al, ^28,29Si, ⁴¹K, ⁵¹V, ⁶¹Ni, ⁶⁵Cu, ^64,67Zn, ⁶⁹Ga, ⁷⁹Br, ⁹²Mo and ⁹³Nb. The cross sections of the ²⁹Si(n, p) ²⁹Al, ⁶⁷Zn(n, p) ⁶⁷Cu, ⁶⁹Ga(n, p) ^69mZn, ⁷⁹Br(n, p) ^79mSe, and ⁶⁹Ga(n, α) ⁶⁶Cu reactions were obtained for the first time in the studied energy range. The d-D neutrons were generated by the deuterium gas target at the Van de Graaff accelerator (KN-VdG) at Nagoya University. All cross section values were determined relative to those of the ¹¹⁵In(n, n′)^115mIn reaction. The activities induced by the low-energy neutrons were corrected. For the corrections, the neutron spectra and mean neutron energies at the irradiation positions were calculated taking into account the energy loss of incident deuterons, the angular differential cross section of the d-D reaction and the solid angle subtended by the sample. The systematics of the (n, p) reactions at the neutron energy of 5.0 MeV in the mass range between 27 and 92 were proposed for the first time. This systematics can predict the cross sections within an accuracy of a factor of 1.6.