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Yamasaki M.,Nuclear Fuel Industries | Unesaki H.,Kyoto University | Yamamoto A.,Nagoya University | Takeda T.,University of Fukui | Mori M.,China Nuclear Power Engineering Co.
Nuclear Technology | Year: 2012

Erbia-credit super high burnup (Er-SHB) fuel offers a means to introduce >5 wt% 235 U enrichment fuel; small amounts of erbia added to all the high-enriched UO 2powder can reduce the initial reactivity to <5 wt% enrichment level. By using this erbia credit, the new fuel can be treated as <5 wt% enriched fuel, and most modifications to the existing facilities and equipment can be avoided. One of the key issues for developing the Er-SHB fuel is to validate the criticality safety analysis tools for this fuel based on a series of experiments using fuel with small amounts of erbia in the entire core. For that purpose, a series of critical experiments have been performed at the Kyoto University Critical Assembly (KUCA). Four critical cores were constructed utilizing two different average enrichments, three different erbia contents, and four different H/U ratios. Numerical analyses have also been performed using several different cross-section libraries, and the results were compared with the measurements from the KUCA experiments. These results confirm the validity of the calculations and the cross-section libraries for determining erbia reactivity. This paper outlines the basic concepts of the Er-SHB fuel, the erbia experiments, and the analyses results. Source

Ozaki T.,Nuclear Fuel Industries | Hibiki T.,Purdue University
Progress in Nuclear Energy | Year: 2015

In view of an important role of a one-dimensional drift-flux correlation in nuclear thermal-hydraulic system analysis codes, several drift-flux correlations such as Lellouche-Zolotar, Chexal-Lellouche, TRAC-BF1 and Ozaki correlations have been reevaluated by rod bundle test data taken in FRIGG and NUPEC test facilities. The mean absolute error of void fraction representing a correlation bias of the Lellouche-Zolotar, Chexal-Lellouche, TRAC-BF1 and Ozaki correlations are, respectively, -1.0, 0.5, -6.3 and -3.3% for the FRIGG test data and 2.0, 2.3, -0.4 and -0.7% for the NUPEC test data. The effects of unheated rods, axial and radial power distributions, large unheated center rod and geometry of a shroud or casing on void fraction are identified. The presence of unheated rods with similar size of other heated rods tends to increase a distribution parameter in a drift-flux correlation, whereas the presence of a large unheated center rod tends to decrease the distribution parameter. The axial and radial power distributions do not have significant effect on void fraction within the tested axial and radial power distribution range. The Ozaki correlation is recommended for predicting void fraction in a BWR core but it is suggested to reduce the distribution parameter in the Ozaki correlation if a large unheated center rod exists in the core. It is indicated that drift-flux correlations developed based on bounded rod bundle test facility data may overestimate the distribution parameter for a PWR core. © 2015 Elsevier Ltd. All rights reserved. Source

Ikeno T.,Nuclear Fuel Industries | Kajishima T.,Osaka University
Nuclear Engineering and Design | Year: 2010

Large eddy simulation (LES) of turbulent flow in a bare rod bundle was performed, and a new concept about the flow structure that enhances heat transport between subchannels was proposed. To investigate the geometrical effect, the LES was performed for three different values of rod diameter over pitch ratio (D/P = 0.7, 0.8, 0.9). The computational domain containing 4 subchannels was large enough to capture large-scale structures wide across subchannels. Lateral flow obtained was unconfined in a subchannel, and some flows indicated a pulsation through the rod gap between subchannels. The gap flow became strong as D/P increased, as existing experimental studies had reported. Turbulence intensity profile in the rod gap suggested that the pulsation was caused by the turbulence energy transferred from the main flow to the wall-tangential direction. This implied that the flow pulsation was an unsteady mode of the secondary flow and arose from the same geometrical effect of turbulence. This implication was supported by the analysis results: two-points correlation functions of fluctuating velocities indicated two length-scales, P-D and D, respectively of cross-sectional and longitudinal motions; turbulence stress in the cross-sectional mean flow contained a non-potential component, which represented energy injection through the unsteady longitudinal fluid motion. © 2008 Elsevier B.V. All rights reserved. Source

Ikeno T.,Nuclear Fuel Industries | Kataoka I.,Osaka University
Nuclear Engineering and Design | Year: 2011

Secondary flow in bubbly turbulent flow in sub-channel was simulated by using an algebraic turbulence stress model. The mass, momentum, turbulence energy and bubble diffusion equations were used as fundamental equation. The basis for these equations was the two-fluid model: the equation of liquid phase was picked up from the equation system theoretically derived for the gas-liquid two-fluid turbulent flow. The fundamental equation was transformed onto a generalized coordinate system fitted to the computational domain in sub-channel. It was discretized for the SIMPLE algorithm using the finite-volume method. The shape of sub-channel causes a distortion of the computational mesh, and orthogonal nature of the mesh is sometimes broken. An iterative method to satisfy a requirement for the contra-variant velocity was introduced to represent accurate symmetric boundary condition. Two-phase flow at a steady state was simulated for different magnitudes of secondary flow and void fraction. The secondary flow enhanced the momentum transport in sub-channel and accelerated the liquid phase in the rod gap. This effect was slightly mitigated when the void fraction increased. The acceleration can contribute to effective cooling in the rod gap. The numerical result implied a phenomenon of industrial interest. This suggested that experimental approach is necessary to validate the numerical model and to identify the phenomenon. © 2011 Elsevier B.V. All rights reserved. Source

Nuclear Fuel Industries | Date: 2010-01-27

The output of a nuclear reactor is increased by a predetermined magnitude, and the neutron beam is measured as time-series data. The temperature of the moderator in the reactor is acquired as time-series data. Time-series data on the reactivity is acquired from the time-series data on the neutron beam by the reverse dynamic characteristic method with respect to a one-point reactor kinetics equation. Time-series data on the fuel temperature of a predetermined average acquired by using the time-series data on the reactor output and a predetermined dynamic characteristic model is acquired. The reactivity feedback contribution component is determined by using the time-series data on the reactivity and the applied reactivity. The Doppler reactivity coefficient is determined by using the time-series data on the average temperature of the moderator in the reactor, the time-series data on the fuel temperature of the predetermined average, the isothermal temperature reactivity coefficient, and the reactivity feedback contribution component.

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