• Journal of Radiation Protection and Research
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• v.41 no.4
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• pp.373-377
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• 2016

Background: Experiments with small electrochemical apparatus were previously carried out for detecting low-energy neutrinos under irradiation of reactor neutrinos and under natural neutrino environment. The experimental result indicated that the output current of reactor-neutrino irradiated detector was appreciably larger than that of natural environmental one. Usual interaction cross-sections of neutrinos are quite small, so that they do not explain the experimental result at all. Materials and Methods: To understand the experimental data, we propose that some biological products may generate AV-type scalar field B0, leading to a large interaction cross-section. The output current generation is ascribed to an electrochemical process that may be assisted by weak interaction phenomena. Dissolved oxygen concentrations in the detector solution were measured in this study, for the purpose of understanding the mechanism of the detector output current generation. Results and Discussion: It was found that the time evolution of experimental output current was mostly reproduced in simulation calculation on the basis of the measured dissolved oxygen concentration. Conclusion: We mostly explained the variation of experimental data by using the electrochemical half-cell analysis model based on the DO concentration that is consistent to the experiment.

• Nuclear Engineering and Technology
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• v.52 no.10
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• pp.2183-2195
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• 2020

A mathematical model for the flowrate and rotation speed of RCP during idling was established. The numerical calculation method and dimensionless method were used to analyze the flow, head, torque and pressure and speed changes under idle conditions. Regularity, using the Q criterion vortex identification judgment method combined with surface flow spectrum morphology analysis to diagnose the vortex dynamic characteristics on RCP blade. On impeller blade, there is two oscillations in the pressure ratio on pressure surface in blade outlet region. The velocity on the suction surface is two times more oscillating than the inlet of blade, and there is an intersection with the velocity ratio curve on pressure surface. On blade of guide vane, the pressure ratio increases along the inlet to outlet direction, and the speed ratio decreases with the increase of idle time. There is a vortex that rotates counterclockwise on the suction surface, and the streamline on the suction surface of blade is subjected to the entrainment and blocking action of the vortex creates a large reverse flow in the main flow region. There are two vortices at the outlet of guide vane suction side and the vortices are in opposite directions.

• Nuclear Engineering and Technology
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• v.53 no.6
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• pp.1846-1857
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• 2021

Ultrasonic nondestructive testing is important for monitoring the structural integrity of dissimilar metal welds (DMWs) in pressure vessels and piping in nuclear power plants. However, there is a low probability of crack detection via inspection of DMWs using ultrasonic waves because the grain structures (grain orientations) of the weld area cause distortion and splitting of ultrasonic beams propagating in anisotropic media. To overcome this issue, the grain orientation should be known, and a precise ultrasonic wave simulation technique in anisotropic media is required to model the distortion and splitting of the waves accurately. In this study, a method for nondestructive prediction of the DMW grain orientations is presented for accurate simulation of ultrasonic wave propagation behavior in the weld area. The ultrasonic wave propagation behavior in anisotropic media is simulated via finite-element analysis when ultrasonic waves propagate in a transversely isotropic material. In addition, a methodology to predict the DMW grain orientation is proposed that employs a simulation technique for ultrasonic wave propagation behavior calculation and an optimization technique. The simulated ultrasonic wave behaviors with the grain orientations predicted via the proposed method demonstrate its usefulness. Moreover, the method can be used to determine the focal law in DMWs.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.326-342
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• 2022

Comparative analyses of a shield building subjected to a large commercial aircraft impact between decoupling method and coupling method are performed in this paper. The decoupling method is applying impact force time-history curves on impact area of the shield building to study impact damage effects on structure. The coupling method is using a model including aircraft and shield building to perform simulation of the entire impact process. Impact force time-history curves of the fuselage, wing and engine and their total impact force time-history curve are obtained by the entire aircraft normally impacting the rigid wall. Taking aircraft structure and impact progress into account some loading areas are determined to perform some comparative analyses between decoupling method and coupling method, the calculation results including displacement, plastic strain of concrete and stress of steel plate in impact area are given. If the loading area is determined unreasonably, it will be difficult to assess impact damage of impact area even though the accurate impact force of each part of aircraft obtained already. The coupling method presented at last in this paper can more reasonably evaluate the dynamic response of the shield building than the decoupling methods used in the current nuclear engineering design.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.357-373
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• 2022

The high-temperature gas-cooled reactor pebble-bed module project is the first commercial Generation-IV NPP(Nuclear Power Plant) in China. A new joint is used for the vertical support of RPV(Reactor Pressure Vessel). The steel corbel is integrally embedded into the reactor-cabin wall through eight asymmetrically arranged pre-stressed high-strength bolts, achieving the different path transmission of shear force and moment. The vertical monotonic loading test of two specimens is conducted. The results show that the failure mode of the joint is bolt fracture. There is no prominent yield stage in the whole loading process. The stress of bolts is linearly distributed along the height of corbel at initial loading. As the load increases, the height of neutral axis of bolts gradually decreases. The upper and lower edges of the wall opening contact the corbel plate to restrict the rotation of the corbel. During the loading, the pre-stress of some bolts decreases. The increase of the pre-stress strength ratio of bolts has no noticeable effect on the structure stiffness, but it reduces the ultimate bearing capacity of the joint. A simplified calculation model for the elastic stage of the joint is established, and the estimation results are in good agreement with the experimental results.

• Nuclear Engineering and Technology
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• v.53 no.7
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• pp.2312-2322
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• 2021

In recent studies, the creep rate of Zircaloy-4, one of the basic property parameters of the nuclear fuel code, has been commonly used with the axial creep model proposed by Rosinger et al. However, in order to calculate the circumferential deformation of the fuel cladding, there is a limitation that a difference occurs depending on the anisotropic coefficients used in deriving the circumferential creep equation by using the axial creep equation. Therefore, in this study, the existing axial creep law and the derived circumferential creep results were analyzed through a circumferential creep test by the internal pressurization method in the isothermal conditions. The circumferential creep deformation was measured through the optical image analysis method, and the results of the experiment were investigated through constructed IDECA (In-situ DEformation Calculation Algorithm based on creep) code. First, preliminary tests were performed in the isotropic β-phase. Subsequently in the anisotropic α-phase, the correlations obtained from a series of circumferential creep tests were compared with the axial creep equation, and optimized anisotropic coefficients were proposed based on the performed circumferential creep results. Finally, the IDECA prediction results using optimized anisotropic coefficients based on creep tests were validated through tube burst tests in transient conditions.

• Nuclear Engineering and Technology
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• v.53 no.1
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• pp.337-343
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• 2021

In this work, self-absorption correction factor related to the variation of the composition and the density of soil samples were evaluated using the p-type HPGe detector. The validated MCNP5 simulation model of this detector was used to evaluate its Full Energy Peak Efficiency (FEPE) under the variation of the composition and the density of the analysed samples. The results indicates that FEPE calculation of low gamma ray is affected by the composition and the density of soil samples. The self-absorption correction factors for different gamma-ray energies which was fitted as a function of FEPEs via density and energy and fitting parameters as polynomial function for the logarithm neper of gamma ray energy help to calculate quickly the detection efficiency of detector. Factor Analysis for the influence of the element composition in analysed samples on the FEPE indicates the FEPE distribution changes from non-metal to metal groups when the gamma ray energy increases from 92 keV to 238 keV. At energies above 238 keV, the FEPE primarily depends only on the metal elements and is significantly affected by aluminium and silicon composition in soil samples.

• Nuclear Engineering and Technology
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• v.22 no.2
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• pp.116-127
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• 1990

The fission gas release model used In the SPEAR-BETA fuel performance code was modified by use of effective thermal conductivity for cracked fuel and by laking Into account axial fission-gas mixing between the fuel-clad gap and the plenum. With use of this modified model the fission gas release was analyzed under various power ramping conditions of P$_{max}$ and $\Delta$.fP. Effective fuel thermal conductivity that accounts for the effect of fuel tracking was used in calculation of the fuel temperature distribution and the Internal gas pressure under power ramping conditions. Mixing and dilution effects due to axial gas flow were also considered in computing the width and the thermal conductivity of the gap. The effect of axial gas flow w3s solved by the Crank-Nicholson method. The finite difference method was used to save running time in the calculation. The present modified fission-gas release model was validated by comparing its predicted results with experimental data from various lamping tests In the literature and calculated results with use of the models used In the SPEAR-BETA and FEMAXI-IV codes. Results obtained with use of the present modified model showed better agreement with experimental data reported in the literature than those results with use of the latter codes. The fuel centerline temperature calculated with introduction of effective thermal conductivity for centerline temperature calculated with Introduction of effective thermal conductivity for cracked fuel was 200 higher fission gas release predicted with use of the modified model was nearly 6% larger on the average than that calculated by use of the unmodified model used in the SPEAR-BETA code.e SPEAR-BETA code.e.

• Journal of the Korea Society for Simulation
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• v.13 no.2
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• pp.13-21
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• 2004

The thermal-hydraulic model ARTS which was based on the RETRAN-3D code adopted in the domestic full-scope power plant simulator which was provided in 1998 by KEPRI. Since ARTS is a generalized code to model the components with control volumes, the smaller time-step size should be used even if converged solution could not get in a single volume. Therefore, dedicated models which do not force to reduce the time-step size are sometimes more suitable in terms of a real-time calculation and robustness. In the case of PRT(Pressurizer Relief Tank) model, it is consist of subcooled water in bottom and non-condensable gas in top. The sparger merged under subcooled water enhances condensation. The complicated thermal-hydraulic phenomena such as condensation, phase separation with existence of non-condensable gas makes difficult to simulate. Therefore, the PRT volume can limit the time-step size if we model it with a general control volume. To prevent the time-step size reduction due to convergence failure for simulating this component, we developed a dedicated model for PRT. The dedicated model was expected to provide substantially more accurate predictions in the analysis of the system transients. The results were resonable in terms of accuracy, real-time simulation, robustness and education of operators, complying with the ANSI/ANS-3.5-1998 simulator software performance criteria and RETRAN-3D results.

• Journal of Energy Engineering
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• v.7 no.2
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• pp.187-193
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• 1998

The turbulent heat transfer to low Prandtl number fluid flow through rod bundles is analyzed using k-$\varepsilon$ two-equation model. For the prediction of the turbulent flow field, an anisotropic eddy viscosity model is used. In the analysis of the temperature field, the effects of various parameters such as geometry, Reynolds and Prandtl numbers are considered. The calculation in made for Prandtl numbers from 0.001 to 0.1 in order to analyze the heat transfer to low Prandtl number fluid such as liquid metals. The numerical results show that for small P/D (Pitch/Diameter) geometries low Prandtl number makes severe changes of the rod surface temperature.