Department of Nuclear Engineering

h i g h l i g h t s Thermo-hydrodynamic properties of water-Al 2 O 3 nanofluid at PWR condition is analyzed. Details of CFD simulation and validation procedure is outlined. Augmented heat transfer capacity of nanofluid is governed by larger pumping power. A new correlation for nanofluid Nusselt number in subchannel geometry is proposed.

This short communication makes use of the principle of singular perturbation to approximate the ordinary differential equation (ODE) of prompt neutron (in the point kinetics model) as an algebraic equation. This approximation is shown to yield a large gain in computational efficiency without compromising any significant accuracy in the numerical simulation of primary coolant system dynamics in a PWR nuclear power plant. The approximate (i.e., singularly perturbed) model has been validated with a numerical solution of the original set of neutron point-kinetic and thermal–hydraulic equations. Both models use variable-step Runge–Kutta numerical integration.

The very high temperature gas-cooled reactor (VHTR) is a uranium-fueled, graphite-moderated, and helium-cooled reactor envisioned as one of the promising future nuclear reactor concepts due to its high efficiency, safety, and variety of applications. In this reactor concept, core bypass flow and cross flow are currently considered as key issues because of their inherent uncertainties and complication of prediction. Recently, the computational fluid dynamics (CFD) method has received a great deal of attention as a method for understanding the flow behavior in the VHTR core, including the bypass and the cross flow. However, validation of this method has not been sufficiently conducted yet based on real experimental data. For this reason, prediction capability of the CFD method was validated in this study by comparing the predictions with the existing multi-hole experimental data obtained by Groehn as a part of the core thermal-hydraulics design study for the NHDD PMR-200. A two-stacked fuel block with wedge-shaped cross gap was simulated for computational domain as the experimental setup. The flow loss coefficients and the velocity distributions of the cross flow from the experiment were compared to the CFD predictions extensively. As a result, good agreements between the CFD predictions and the experimental results were observed, confirming prediction capability of the CFD method for the complicated cross flow in the VHTR core. Furthermore, the velocity distributions and pressure distribution in the cross gap were investigated to identify the characteristics of the cross flow.

This paper introduces the results of numerical simulations on flow fields and relevant heat transfer in the pebble bed reactor (PBR) core. In the core, since the coolant passes a highly complicated random flow path with a high Reynolds number, an appropriate treatment of the turbulence is required. A set of simple experiments for the flow over a circular cylinder with heat transfer was conducted to finally select the large eddy simulation (LES) and k-! model among the considering Reynolds-averaged Navier-Stokes (RANS) models for PBR application. Using these models, the PBR cores, whose geometries were simplified to the body-centered cubical (BCC) and face-centered cubical (FCC) structures, were simulated. A larger pressure drop, a more random flow field, a higher vorticity magnitude and a higher temperature at the local hot spots on the pebble surface were found in the results of the LES than in those of RANS for both geometries. In cases of the LES, the flow structures were resolved up to the grid scales. Irregular distributions of the flow and local heat transfer were found in the BCC core, while relatively regular distributions for the FCC core. The turbulent nature of the coolant flow in the pebble core evidently affected the fuel surface temperature distribution.

This paper introduces the results of numerical simulations on flow fields and relevant heat transfer in the pebble bed reactor (PBR) core. In the core, since the coolant passes a highly complicated random flow path with a high Reynolds number, an appropriate treatment of the turbulence is required. A set of simple experiments for the flow over a circular cylinder with heat transfer was conducted to finally select the large eddy simulation (LES) and k-ω model among the considering Reynolds-averaged Navier-Stokes (RANS) models for PBR application. Using these models, the PBR cores, whose geometries were simplified to the body-centered cubical (BCC) and face-centered cubical (FCC) structures, were simulated. A larger pressure drop, a more random flow field, a higher vorticity magnitude and a higher temperature at the local hot spots on the pebble surface were found in the results of the LES than in those of RANS for both geometries. In cases of the LES, the flow structures were resolved up to the grid scales. Irregular distributions of the flow and local heat transfer were found in the BCC core, while relatively regular distributions for the FCC core. The turbulent nature of the coolant flow in the pebble core evidently affected the fuel surface temperature distribution.

VHTR is a gas-cooled, graphite-moderated reactor that was developed for the production of hydrogen and process heat. In the prismatic VHTR core with hexagonal graphite blocks, the bypass gap, which is an interstitial gap between the hexagonal blocks, inevitably forms due to the installation tolerance and irradiation shrinkage of the core blocks. Core bypass flow is the ineffective coolant flow that does not pass through the coolant holes within the fuel block. The core-bypass flow distribution varies according to the fuel cycle because the graphite blocks are deformed by irradiation and heating. Since the bypass flow reduces the core thermal margin, it should be evaluated precisely to guarantee reactor safety. In this regard, experimental and computational studies were carried out to evaluate the core-bypass flow distribution. A multi-block air test facility was designed and constructed in Seoul National University to measure the flow and pressure distributions according to the block combination and cross-flow gap size. The effects of the cross-flow phenomenon on the flow and pressure distribution in the core were investigated. The experimental data were used to validate a CFD model for the analysis of bypass flow characteristics in detail.

Uncertainty analysis in Modelling (UAM) for Design, Operation and Safety Analysis of Sodium-cooled Fast Reactors (SFRs) has been formed by OECD/NEA to assess the effect of nuclear data uncertainties on parameters of interest in SFR analysis. In this paper, sub-exercises of a medium 1000 MWth metallic core (MET-1000) and a large 3600 MWth oxide core (MOX-3600) are tested by a Monte Carlo code MCS to perform uncertainty analysis. Classical perturbation theory and generalized perturbation theory are used to calculate sensitivity coefficients. Uncertainty is calculated by multiplying the sensitivity coefficients and relative covariance matrix from ENDF/B-VII.1 library.

The theoretical aspects behind the reactor depletion capability of the Monte Carlo code MCS developed at the Ulsan National Institute of Science and Technology (UNIST) and practical results of this depletion feature for a Material-Testing Reactor (MTR) with plate-type fuel are described in this paper. A verification of MCS results is first performed against MCNP6 to confirm the suitability of MCS for the criticality and depletion analysis of the MTR. Then, the dependence of the effective neutron multiplication factor to the number of axial and radial depletion cells adopted in the fuel plates is performed with MCS in order to determine the minimum spatial segmentation of the fuel plates. Monte Carlo depletion results with 37,800 depletion cells are provided by MCS within acceptable calculation time and memory usage. The results show that at least 7 axial meshes per fuel plate are required to reach the same precision as the reference calculation whereas no significant differences are observed when modeling 1 or 10 radial meshes per fuel plate. This study demonstrates that MCS can address the need for Monte Carlo codes capable of providing reference solutions to complex reactor depletion problems with refined meshes for fuel management and research reactor applications.

MCS is a Monte Carlo (MC) neutron transport simulation code which has been developed exclusively since 2013 by the COmputational Reactor physics and Experiment laboratory (CORE) of Ulsan National Institute of Science and Technology (UNIST) for the neutronics design of nuclear reactors. To validate the capabilities of UNIST MCS, the OPR-1000 is modeled and the 3D whole-core depletion calculation is coupled with various feedback options, one dimensional thermal hydraulics (TH1D), depletion, Critical Boron Concentration (CBC), Equilibrium Xenon (Eq-Xe), and OpenW. Then, the calculated results of the axial and radial power distributions at Middle Of Cycle (MOC) and End Of Cycle (EOC) are compared with measured data. In addition, the result of the CBC calculation in MCS is compared with the measurement data and Nuclear Design Report (NDR) data. Overall, this study shows good agreement between MCS and the measured data. Therefore, good performance of the MCS modelling capabilities for OPR...

Current interest in advanced nuclear energy and molten salt reactor (MSR) concepts characterized by its use of the liquid-fuel salts serving as a fuel and as a coolant has enhanced interest in developing new core design and building the tools necessary to analyze the reactor core system. Seaborg Technologies is actively developing a design for its compact molten salt reactor (CMSR)[1], a liquid-fuel MSR with a thermal power output of 250 MWth. The liquid-fuel of the reactor provides important benefits over the traditional solidfuel reactors with passive safety and fuel utilization. However, during the fuel circulation, the position of delayed neutron precursors continuously changes both in the core and in the external loop, and the fission products (FPs) are extracted by an online fuel reprocessing unit. The online fuel reprocessing limits to most of the simulation tools used with the solid-fuel in the conventional reactor. One exception is the Monte Carlo code Serpent 2 [2] which h...

Jinsu Park a, Jiwon Choe a, Hanjoo Kim a, Sooyoung Choi a, Alexey Cherezov a, Jaerim Jang a, Dongmin Yun a, Nhan Nguyen Trong Mai a, Vutheam Dos a, Deokjung Lee a*, Ho Cheol Shin b, Hwan Soo Lee b aDepartment of Nuclear Engineering, Ulsan National Institute of Science and Technology, 50 UNIST-gil, Ulsan, 44919 bCore and Fuel Analysis Group, Korea Hydro & Nuclear Power Central Research Institute, Daejeon, 34101 *Corresponding author: deokjung@unist.ac.kr

Department of Nuclear EngineeringWorldwide demand for high-fidelity simulation tools for large scale power reactor analysis resulted in the coupling of neutronics, thermal hydraulics, and fuel mechanics in nuclear reactor cores. The Computational Reactor Physics and Experiment (CORE) laboratory at Ulsan National Institute of Science and Technology (UNIST) has developed an inhouse Monte Carlo code MCS coupled with multi-physics (MP) tools such as a one-dimensional (1D) single-phase closed-channel thermal-hydraulic code (TH1D), a sub-channel two-phase thermal-hydraulic code (CTF) and a fuel performance code (FRAPCON) to provide thermal-hydraulic and fuel behavior feedbacks for realistic applications. This thesis describes the 3D whole-core analysis with pin-wise resolution in which neutron transport, depletion, thermal-hydraulic, and fuel behavior calculations are performed using the coupled MCS/TH1D, MCS/CTF, and MCS/FRAPCON tools. The OPR-1000 PWR core operated for 2 consecutive cyc...

Uncertainty analysis in Modelling (UAM) for Design, Operation and Safety Analysis of Sodium-cooled Fast Reactors (SFRs) has been formed by OECD/NEA to assess the effect of nuclear data uncertainties on parameters of interest in SFR analysis. In this paper, subexercises of a medium 1000 MWth metallic core (MET-1000) and a large 3600 MWth oxide core (MOX-3600) are tested by a Monte Carlo code MCS to perform uncertainty analysis. Classical perturbation theory and generalized perturbation theory are used to calculate sensitivity coefficients. Uncertainty is calculated by multiplying the sensitivity coefficients and relative covariance matrix from ENDF/B-VII.1 library.

Recent researches have become more interested in the feasibility of using Monte Carlo (MC) code to generate multi-group (MG) cross sections (XSs) for fast reactor analysis using nodal diffusion codes. The current study, therefore, presents a brief methodology for MG XSs generation by the in-house UNIST MC code MCS, which can be compatibly utilized in nodal diffusion codes, PARCS and RAST-K. The applicability of the methodology is quantified on the sodium fast reactor (SFR) ABR-1000 design with a metallic fuel from the OECD/NEA SRF benchmark. The few-group XSs generated by MCS with a two-dimensional (2D) fuel assembly geometry are well consistent with those of SERPENT 2. Furthermore, the simulation of beginning-of-cycle (BOC) steady-state three-dimensional (3D) whole-core problem with PARCS and RAST-K is conducted using the generated 24-group XSs by MCS. The nodal diffusion solutions, including the core keff, power profiles and various of reactivity parameters, are compared to reference whole-core results obtained by MC code MCS. Overall, the codeto-code comparison indicates a reasonable agreement between deterministic and stochastic codes, with the difference in keff less than 100 pcm and the root-mean-square (RMS) error in assembly power less than 1.15%. Therefore, it is successfully demonstrated that the employment of the MG XSs generation by MCS for nodal diffusion codes is feasible to accurately perform analyses for fast reactors.

The theoretical aspects behind the reactor depletion capability of the Monte Carlo code MCS developed at the Ulsan National Institute of Science and Technology (UNIST) and practical results of this depletion feature for a Material-Testing Reactor (MTR) with plate-type fuel are described in this paper. A verification of MCS results is first performed against MCNP6 to confirm the suitability of MCS for the criticality and depletion analysis of the MTR. Then, the dependence of the effective neutron multiplication factor to the number of axial and radial depletion cells adopted in the fuel plates is performed with MCS in order to determine the minimum spatial segmentation of the fuel plates. Monte Carlo depletion results with 37,800 depletion cells are provided by MCS within acceptable calculation time and memory usage. The results show that at least 7 axial meshes per fuel plate are required to reach the same precision as the reference calculation whereas no significant differences are observed when modeling 1 or 10 radial meshes per fuel plate. This study demonstrates that MCS can address the need for Monte Carlo codes capable of providing reference solutions to complex reactor depletion problems with refined meshes for fuel management and research reactor applications.

This study presents the first application of the advanced Rossi-alpha method (theoretically introduced by Kong et al., 2014) on the reactivity measurements in a research reactor: detector count signals at the Kyoto University Critical Assembly (KUCA) facility. The detector signals in the KUCA A-type core are analyzed by three subcriticality measurement methods: (1) Feynman-alpha (F-a) method, (2) Rossi-alpha (R-a) method, and (3) advanced Rossi-alpha (advanced R-a) method. Four cases are analyzed for two different subcritical states of the core and two different neutron source locations. Two different negative reactivity q values are obtained by the measurements of control rod worth and regarded as the reference reactivity values, comparing the results by the four methods. The F-a shows reactivity errors ranging between 7.1 and 7.3% due to its use of variance-to-mean ratios of detector count signals, which are not very sensitive to neutron background noise. However, the fitting uncertainties associated to the F-a results are large, ranging between 5.4 and 12.8% at one standard deviation. The R-a shows small fitting uncertainties ranging between 2.8 and 3.8%, although reactivity errors are in the range of 3.5-26.5% due to the neutron background noise. Finally, the advanced R-a that explicitly models the neutron background noise contrary to the previous methods shows the reactivity errors in the range of 1.0-11.8%, and provides the lowest uncertainties of the measured q in the range of 0.4-0.9%. In conclusion, among the four methods applied to the reactivity measurements at KUCA, the advanced Ra reveals the best accuracy with the lowest uncertainties.

In cold pulse experiments in JTEXT, not only are rapid electron temperature increases in the core observed, but also steep rises in the inner density are found. Moreover, some evidence of acceleration of the core toroidal rotation is also observed during the nonlocal transport process of electron temperature. These new findings of cold pulse experiments in JTEXT suggest that turbulence spreading is a possible mechanism for the nonlocal transport dynamics.