• Nuclear Engineering and Technology
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• v.49 no.1
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• pp.29-42
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• 2017

The purpose of the present study is the presentation of the appropriate element and shape function in the solution of the neutron diffusion equation in two-dimensional (2D) geometries. To this end, the multigroup neutron diffusion equation is solved using the Galerkin finite element method in both rectangular and hexagonal reactor cores. The spatial discretization of the equation is performed using unstructured triangular and quadrilateral finite elements. Calculations are performed using both linear and quadratic approximations of shape function in the Galerkin finite element method, based on which results are compared. Using the power iteration method, the neutron flux distributions with the corresponding eigenvalue are obtained. The results are then validated against the valid results for IAEA-2D and BIBLIS-2D benchmark problems. To investigate the dependency of the results to the type and number of the elements, and shape function order, a sensitivity analysis of the calculations to the mentioned parameters is performed. It is shown that the triangular elements and second order of the shape function in each element give the best results in comparison to the other states.

• Nuclear Engineering and Technology
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• v.53 no.4
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• pp.1311-1317
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• 2021

In this study, an airborne alpha detection system, which consists of a passivated implanted planar silicon (PIPS) detector and an air filter, was developed. A collimator applied to the alpha detection system showed an enhancement in resolution and a degradation in detection efficiency. The resolution and detection efficiency were compared and analyzed to evaluate the performance of the collimator. Thus, the resolution was found to be more important than the efficiency as a determining factor of the detection system performance, from the viewpoint of radionuclide identification. The performance was evaluated on three properties of the collimator: hole shape, hole length, and the ratio between the hole and frame pitches. From the hole shape performance evaluation, a hexagonal collimator showed the highest resolution. Further, the collimator with a hole pitch of 14 mm was found to have the highest resolution while that with a frame pitch of 4-6 mm (i.e., 1.2-1.4 times longer than the hole pitch) showed the highest resolution.

• Nuclear Engineering and Technology
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• v.44 no.1
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• pp.53-60
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• 2012

The aim of this study was to explore the spatial and energy resolutions of a PET scanner that we have recently developed. The scanner, which consists of six detector modules with 1-layer LGSO crystals, has a hexagonal configuration with a faceto- face distance of 86.4 mm between two opposite PET modules; such properties facilitate the imaging of small animals. A $^{22}Na$ point source was employed to estimate horizontal and vertical spatial resolutions. To assess the energy resolution, a uniform $^{18}F$ cylindrical phantom was scanned. A software-based spectrum analysis of list-mode data was used to assign a local energy window centered on the photopeak position for every single crystal. For the image reconstruction, an ML-EM algorithm was used. The spatial resolutions at the center of the scanner were 0.99 mm in the horizontal direction and 1.13 mm in the vertical direction. The energy resolution averaged over each PMT ranged from 13.3%-14.3%, which gave an average value of 13.8%. These results show that this simple system is promising for small animal imaging with excellent spatial and energy resolutions.

• Nuclear Engineering and Technology
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• v.47 no.5
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• pp.523-533
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• 2015

Three-dimensional flow phenomena in a wire-wrapped 37-pin fuel assembly mock-up of a Japanese loop-type sodium-cooled fast reactor, Monju, were investigated with a numerical analysis using a general-purpose commercial computational fluid dynamics code, CFX. Complicated and vortical flow phenomena in the wire-wrapped 37-pin fuel assembly were captured by a Reynolds-averaged Navier-Stokes flow simulation using a shear stress transport turbulence model. The main purpose of the current study is to understand the three-dimensional complex flow phenomena in a wire-wrapped fuel assembly to support the license issue for the core design. Computational fluid dynamics results show good agreement with friction factor correlation models. The secondary flow in the corner and edge subchannels is much stronger than that in an interior subchannel. The axial velocity averaged in the corner and edge subchannels is higher than that averaged in the interior subchannels. Three-dimensional multiscale vortex structures start to be formed by an interaction between secondary flows around each wire-wrapped pin. Behavior of the large-scale vortex structures in the corner and edge subchannels is closely related to the relative position between the hexagonal duct wall and the helically wrapped wire spacer. The small-scale vortex is axially developed in the interior subchannels. Furthermore, a driving force on each wire spacer surface is closely related to the relative position between the hexagonal duct wall and the wire spacer.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.3902-3909
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• 2021

SACOS series of subchannel analysis codes have been developed by XJTU-NuTheL for many years and are being used for the thermal-hydraulic safety analysis of various reactor cores. To achieve fine whole core pin-level analysis, the input preprocessing and parallel capabilities of the code have been developed in this study. Preprocessing is suitable for modeling rectangular and hexagonal assemblies with less error-prone input; parallelization is established based on the domain decomposition method with the hybrid of MPI and OpenMP. For domain decomposition, a more flexible method has been proposed which can determine the appropriate task division of the core domain according to the number of processors of the server. By performing the calculation time evaluation for the several PWR assembly problems, the code parallelization has been successfully verified with different number of processors. Subsequent analysis results for rectangular- and hexagonal-assembly core imply that the code can be used to model and perform pin-level core safety analysis with acceptable computational efficiency.

• Nuclear Engineering and Technology
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• v.49 no.7
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• pp.1379-1387
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• 2017

This paper presents a comparison of homogenization techniques implemented in the APOLLO3 platform for transport core calculations: standard scalar flux weighting and new flux-moment homogenization, in different combinations with (or without) leakage models. Besides the historical B1-homogeneous model, a new B-heterogeneous one has indeed been implemented recently in the two/three-dimensional-transport solver using the method of characteristics. First analyses have been performed on a very simple Sodium Fast Reactor core with a regular hexagonal lattice. They show that using the heterogeneous leakage model in association with flux-moment homogenization strongly improves the prediction of $k_{eff}$ and void reactivity effects. These good results are confirmed when the application is done to the fissile assemblies of the more complex CFV (Low Void Effect) core of the ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) project of sodium-cooled fast breeder reactor (Generation IV).

• Nuclear Engineering and Technology
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• v.48 no.1
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• pp.43-54
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• 2016

In the present paper, development of the three-dimensional (3D) computational code based on Galerkin finite element method (GFEM) for solving the multigroup forward/adjoint diffusion equation in both rectangular and hexagonal geometries is reported. Linear approximation of shape functions in the GFEM with unstructured tetrahedron elements is used in the calculation. Both criticality and fixed source calculations may be performed using the developed GFEM-3D computational code. An acceptable level of accuracy at a low computational cost is the main advantage of applying the unstructured tetrahedron elements. The unstructured tetrahedron elements generated with Gambit software are used in the GFEM-3D computational code through a developed interface. The forward/adjoint multiplication factor, forward/adjoint flux distribution, and power distribution in the reactor core are calculated using the power iteration method. Criticality calculations are benchmarked against the valid solution of the neutron diffusion equation for International Atomic Energy Agency (IAEA)-3D and Water-Water Energetic Reactor (VVER)-1000 reactor cores. In addition, validation of the calculations against the $P_1$ approximation of the transport theory is investigated in relation to the liquid metal fast breeder reactor benchmark problem. The neutron fixed source calculations are benchmarked through a comparison with the results obtained from similar computational codes. Finally, an analysis of the sensitivity of calculations to the number of elements is performed.

• Nuclear Engineering and Technology
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• v.50 no.3
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• pp.340-355
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• 2018

The methods and performance of a fast reactor multigroup cross section (XS) generation code EXUS-F are described that is capable of directly processing Evaluated Nuclear Data File format nuclear data files. RECONR of NJOY is used to generate pointwise XS data, and Doppler broadening is incorporated by the Gauss-Hermite quadrature method. The self-shielding effect is incorporated in the ultrafine group XSs in the resolved and unresolved resonance ranges. Functions to generate scattering transfer matrices and fission spectrum matrices are realized. The extended transport approximation is used in zero-dimensional calculations, whereas the collision probability method and the method of characteristics are used for one-dimensional cylindrical geometry and two-dimensional hexagonal geometry problems, respectively. Verification calculations are performed first for various homogeneous mixtures and cylindrical problems. It is confirmed that the spectrum calculations and the corresponding multigroup XS generations are performed adequately in that the reactivity errors are less than 50 pcm with the McCARD Monte Carlo solutions. The nTRACER core calculations are performed with the EXUS-F-generated 47 group XSs for the two-dimensional Advanced Burner Reactor 1000 benchmark problem. The reactivity error of 160 pcm and the root mean square error of the pin powers of 0.7% indicate that EXUF-F generates properly the broad-group XSs.

• Nuclear Engineering and Technology
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• v.47 no.7
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• pp.814-826
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• 2015

The most important advantage of nanoparticles is the increased thermal conductivity coefficient and convection heat transfer coefficient so that, as a result of using a 1.5% volume concentration of nanoparticles, the thermal conductivity coefficient would increase by about twice. In this paper, the effects of a nanofluid ($TiO_2$/water) on heat transfer characteristics such as the thermal conductivity coefficient, heat transfer coefficient, fuel clad, and fuel center temperatures in a VVER-1000 nuclear reactor are investigated. To this end, the cell equivalent of a fuel rod and its surrounding coolant fluid were obtained in the hexagonal fuel assembly of a VVER-1000 reactor. Then, a fuel rod was simulated in the hot channel using Computational Fluid Dynamics (CFD) simulation codes and thermohydraulic calculations (maximum fuel temperature, fluid outlet, Minimum Departure from Nucleate Boiling Ratio (MDNBR), etc.) were performed and compared with a VVER-1000 reactor without nanoparticles. One of the most important results of the analysis was that heat transfer and the thermal conductivity coefficient increased, and usage of the nanofluid reduced MDNBR.

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.13-30
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• 2019

This paper presents the recent improvements in the DeCART code for HTGR analysis. A new 190-group DeCART cross-section library based on ENDF/B-VII.0 was generated using the KAERI library processing system for HTGR. Two methods for the eigen-mode adjoint flux calculation were implemented. An azimuthal angle discretization method based on the Gaussian quadrature was implemented to reduce the error from the azimuthal angle discretization. A two-level parallelization using MPI and OpenMP was adopted for massive parallel computations. A quadratic depletion solver was implemented to reduce the error involved in the Gd depletion. A module to generate equivalent group constants was implemented for the nodal codes. The capabilities of the DeCART code were improved for geometry handling including an approximate treatment of a cylindrical outer boundary, an explicit border model, the R-G-B checker-board model, and a super-cell model for a hexagonal geometry. The newly improved and implemented functionalities were verified against various numerical benchmarks such as OECD/MHTGR-350 benchmark phase III problems, two-dimensional high temperature gas cooled reactor benchmark problems derived from the MHTGR-350 reference design, and numerical benchmark problems based on the compact nuclear power source experiment by comparing the DeCART solutions with the Monte-Carlo reference solutions obtained using the McCARD code.