• Nuclear Engineering and Technology
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• 제51권8호
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• pp.1886-1896
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• 2019

Modern small modular nuclear reactors can be built on a barge in ocean, therefore, their flow characteristics depend upon the ocean motions. In the present research, effect of rolling motion on flow and friction characteristics of laminar flow through vertical and horizontal narrow channels has been studied. A computer code has been developed using MPS method for two-dimensional Navier-Stokes equations with rolling motion force incorporated. Numerical results have been validated with the literature and have been found in good agreement. It has been found that the impact of rolling motions on flow characteristics weakens with increase in flow rate and fluid viscosity. For vertical narrow channels, the time averaged friction coefficient for vertical channels differed from steady friction coefficient. Furthermore, increasing the horizontal distance from rolling pivot enhanced the flow fluctuations but these stayed relatively unaffected by change in vertical distance of channel from the rolling axis. For horizontal narrow channels, the flow fluctuations were found to be sinusoidal in nature and their magnitude was found to be dependent mainly upon gravity fluctuations caused by rolling.

• Nuclear Engineering and Technology
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• 제52권5호
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• pp.878-888
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• 2020

This paper presents the validation of UNIST in-house Monte Carlo code MCS used for the high-fidelity simulation of commercial pressurized water reactors (PWRs). Its focus is on the accurate, spatially detailed neutronic analyses of startup physics tests for the initial core of the Watts Bar Nuclear 1 reactor, which is a vital step in evaluating core phenomena in an operating nuclear power reactor. The MCS solutions for the Consortium for Advanced Simulation of Light Water Reactors (CASL) Virtual Environment for Reactor Applications (VERA) core physics benchmark progression problems 1 to 5 were verified with KENO-VI and Serpent 2 solutions for geometries ranging from a single-pin cell to a full core. MCS was also validated by comparing with results of reactor zero-power physics tests in a full-core simulation. MCS exhibits an excellent consistency against the measured data with a bias of ±3 pcm at the initial criticality whole-core problem. Furthermore, MCS solutions for rod worth are consistent with measured data, and reasonable agreement is obtained for the isothermal temperature coefficient and soluble boron worth. This favorable comparison with measured parameters exhibited by MCS continues to broaden its validation basis. These results provide confidence in MCS's capability in high-fidelity calculations for practical PWR cores.

• Nuclear Engineering and Technology
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• 제53권4호
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• pp.1079-1087
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• 2021

NTP-ERSN is a package developed for solving the multigroup form of the discrete ordinates, characteristics and collision probability of the Boltzmann transport equation in one-dimensional cartesian geometry, by combining pin cells. In this work, C5G7 MOX benchmark is used to verify the accuracy and efficiency of NTP-ERSN package, by treating reactor core problems without spatial homogenization. This benchmark requires solutions in the form of normalized pin powers as well as the vectors and the eigenvalue. All NTP-ERSN simulations are carried out with appropriate spatial and angular approximations. A good agreement between NTP-ERSN results with those obtained with OpenMC calculation code for seven energy groups. In addition, our studies about angular and mesh refinements are carried out to produce better quality solution. Moreover, NTP-ERSN GUI has also been updated and adapted to python 3 programming language.

• Nuclear Engineering and Technology
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• 제53권6호
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• pp.1810-1820
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• 2021

Study of single bubble behavior under rolling motions can prove useful for fundamental understanding of flow field inside the modern small modular nuclear reactors. The objective of the present study is to simulate the influence of rolling conditions on single rising bubble in a liquid using multiphase Moving Particle Semi-implicit (MPS) method. Rolling force term was added to 2D Navier-Stokes equations and a computer program was written using C language employing OpenACC to port the code to GPU. Computational results obtained were found to be in good agreement with the results available in literature. The impact of rolling parameters on trajectory and velocity of the rising bubble has been studied. It has been found that bubble rise velocity increases with rolling amplitude due to modification of flow field around the bubble. It has also been concluded that the oscillations of free surface, caused by rolling, influence the bubble trajectory. Furthermore, it has been discovered that smaller vessel width reduces the impact of rolling motions on the rising bubble. The effect of liquid viscosity on bubble rising under rolling was also investigated and it was found that effects of rolling became more pronounced with the increase of liquid viscosity.

• Nuclear Engineering and Technology
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• 제55권9호
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• pp.3213-3228
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• 2023

As an important part of the digital reactor, the pin-by-pin wise fine coupling calculation is a research hotspot in the field of nuclear engineering in recent years. It provides more precise and realistic simulation results for reactor design, operation and safety evaluation. CORCA-K a nodal code is redeveloped as a robust pin-by-pin wise neutronics and thermal-hydraulic coupled calculation code for pressurized water reactor (PWR) core. The nodal green's function method (NGFM) is used to solve the three-dimensional space-time neutron dynamics equation, and the single-phase single channel model and one-dimensional heat conduction model are used to solve the fluid field and fuel temperature field. The mesh scale of reactor core simulation is raised from the nodal-wise to the pin-wise. It is verified by two benchmarks: NEACRP 3D PWR and PWR MOX/UO₂. The results show that: 1) the pin-by-pin wise coupling calculation system has good accuracy and can accurately simulate the key parameters in steady-state and transient coupling conditions, which is in good agreement with the reference results; 2) Compared with the nodal-wise coupling calculation, the pin-by-pin wise coupling calculation improves the fuel peak temperature, the range of power distribution is expanded, and the lower limit is reduced more.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.933-940
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• 2024

A macroscopic cross section generation model was developed for the conceptual horizontal, compact high temperature gas reactor (HC-HTGR). Because there are many sources of spectral effects in the design and analysis of the core, conventional LWR methods have limitations for accurate simulation of the HC-HTGR using a neutron diffusion core neutronics simulator. Several super-cell model configurations were investigated to consider the spectral effect of neighboring cells. A new history variable was introduced for the existing library format to more accurately account for the history effect from neighboring nodes and reactivity control drums. The macroscopic cross section library was validated through comparison with cross sections generated using full core Monte Carlo models and single cell cross section for both 3D core steady-state problems and 2D and 3D depletion problems. Core calculations were then performed with the AGREE HTR neutronics and thermal-fluid core simulator using super-cell cross sections. With the new history variable, the super-cell cross sections were in good agreement with the full core cross sections even for problems with significant spectrum change during fuel shuffling and depletion.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2006년도 춘계학술대회논문집
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• pp.1306-1311
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• 2006

The root cause of abnormal sound from a heat exchanger of condensate water system in a nuclear power plant is investigated by using the impact signal-processing methodology based on the Hertz theory. The predicted results for the location of impact force and the loose part size meet good agreement with the identified materials during the overhaul period in the plant. Nuclear power plants have experienced several loose parts and the frequency of the loose part will be increased along the aging of the plants. So, this analysis methodology for the impact signal will be widely utilized for the primary and secondary side of the nuclear power plant.

• Nuclear Engineering and Technology
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• 제41권7호
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• pp.893-906
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• 2009

The present work presents the development of a Large Eddy Simulation (LES) methodology viable for complex geometries and suitable for the simulation of rod-bundles. The use of LES and Direct Numerical Simulation (DNS) allows for a deeper analysis of the flow field and the use of stochastical tools in order to obtain additional insight into rod-bundle hydrodynamics. Moreover, traditional steady-state CFD simulations fail to accurately predict distributions of velocity and temperature in rod-bundles when the pitch (P) to diameter (D) ratio P/D is smaller than 1.1 for triangular lattices of cylindrical pins. This deficiency is considered to be due to the failure to predict large-scale coherent structures in the region of the gap. The main features of the code include multi-block capability and the use of the fractional step algorithm. As a Sub-Grid-Scale (SGS) model, a Dynamic Smagorinsky model has been used. The code has been tested on plane channel flow and the flow in annular ducts. The results are in excellent agreement with experiments and previous calculations.

• Nuclear Engineering and Technology
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• 제46권6호
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• pp.863-868
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• 2014

The evaluation of thermal properties of actinide oxide fuels is a problem of high importance for the development of new generation reactors. In the present study, an expression obtained for n-dimensional Debye functions is used to derive a simple analytical expression for the specific heat capacity of nuclear fuels. To test the validity and reliability of this expression, the analytical expression is applied to $UO_2$, $NpO_2$, $ThO_2$, and $PuO_2$. It is seen that the formula was in agreement with the experimental and theoretical results reported in the literature.

• Nuclear Engineering and Technology
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• 제50권1호
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• pp.145-150
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• 2018

A new methodology, the crack-spallation model, has been developed to analyze gas-solid reactions dominated by crack growth inside of the solid reactant and spallation phenomena. The new model physically represents three processes of the reaction progress: (1) diffusion of gas reactant through pores; (2) growth of product particle in pores; and (3) crack and spallation of solid reactant. The validation of this method has been conducted by comparison of results obtained in an experiment for oxidation of $UO_2$ and the shrinking core model. The reaction progress evaluated by the crack-spallation model shows better agreement with the experimental data than that evaluated by the shrinking core model. To understand the trigger point during the reaction progress, a detailed analysis has been conducted. A parametric study also has been performed to determine mass diffusivities of the gas reactant and volume increase constants of the product particles. This method can be appropriately applied to the gas-solid reaction based on the crack and spallation phenomena such as the voloxidation process.