scholarly journals Generating Hexahedral Mesh for Wire-wrapped Fuel Assembly With RBF Mesh Deformation Method

2021 ◽  
Vol 8 ◽  
Author(s):  
X. A. Wang ◽  
Dalin Zhang ◽  
Mingjun Wang ◽  
Yapeng Liu ◽  
Lei Zhou ◽  
...  
Keyword(s):  
Nuclear Reactor ◽  
Fluid Dynamic ◽  
Rapid Development ◽  
Three Dimensional ◽  
Reactor Core ◽  
Hexahedral Mesh ◽  
Deformation Method ◽  
Fuel Assemblies ◽  
Polyhedral Cells ◽  
Cfd Method

Fuel assemblies with wire spacer are widely used in Generation IV liquid nuclear reactors. With the rapid development of computational power, the Computational Fluid Dynamic (CFD) method is becoming an effective tool to investigate the detailed three-dimensional thermal hydraulic characteristics in wire-wrapped fuel assemblies. Due to the complexity of geometry, most of the published researches are performed with large number tetrahedron or polyhedral cells. The simulation is quite time-consuming and is generally limited to assemblies with small number of fuel pins. In this paper, a hexahedron meshing strategy is developed based on the Radial Basis Function (RBF) theory in present paper. This strategy would be beneficial for the modeling for the wire-wrapped fuel assemblies in real nuclear reactor core with large number of fuel pins. To validate this strategy, two experiments are simulated and detailed flow parameter distributions within the bundle, including the pressure distribution and the temperature distribution, have been compared. Good agreements have been achieved between the simulation results and the experimental results.

scholarly journals THE DEVELOPMENT OF A THREE-DIMENSIONAL MODEL OF WWER-1000 CORE USING THE MONTE CARLO SERPENT CODE FOR NEUTRON-PHYSICAL MODELING

2019 ◽  
pp. 58-61
Author(s):  
V.I. Gulik ◽  
O.R. Trofymenko ◽  
V.V. Galchenko ◽  
D.V. Budik
Keyword(s):  
Monte Carlo ◽  
3D Modeling ◽  
Physical Modeling ◽  
Three Dimensional ◽  
Reactor Core ◽  
Dimensional Model ◽  
Three Dimensional Model ◽  
Fuel Assemblies

The article presents the use of the new Monte Carlo Serpent code for 3D modeling of the WWER-1000 reactor core. Core models for the first loading of RNPP4 and the 28th loading of SUNPP3, the fuel assemblies’ models of different manufacturers were developed and presented. Considerable attention was paid to the detailed modeling of the upper, lower and side reflectors. Validation calculations of the Monte Carlo Serpent code for the WWER-1000 reactor were performed on the basis of the first RNPP4 loading. For the 28th loading of SUNPP3, albedo coefficients for radial and axial reflectors were obtained.

Numerical Simulated Method on Wind Environmental of Outdoor

2011 ◽  
Vol 250-253 ◽  
pp. 3815-3821
Author(s):  
Can Li ◽  
Tian Fu Deng
Keyword(s):  
Numerical Simulation ◽  
Fluid Dynamic ◽  
Three Dimensional ◽  
Outdoor Air ◽  
Wind Environment ◽  
Outdoor Air Quality ◽  
Turbulent Airflow ◽  
Cfd Method ◽  
Velocity Pressure ◽  
Improve Planning

Outdoor wind environment play an important role at planning an estate. The method of simulating the wind environment of the district was conducted by two cases that had been tested and verified. The steady-state three-dimensional turbulent airflow fields were analyzed by computational fluid dynamic (CFD) method for the two cases. The distribution of velocity, pressure and turbulivity were presented. The detailed schemes on simplified physical model, mesh division, solving and boundary conditions defining were presented. Results shows that numerical simulation helps predict the details of the outdoor wind environment of estate, and it helps evaluate outdoor air quality in all its aspects or improve planning.

scholarly journals Powering Down

2003 ◽  
Vol 125 (04) ◽  
pp. 46-48
Author(s):  
Harry Hutchinson
Keyword(s):  
Heat Transfer ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
The United States ◽  
Nuclear Industry ◽  
Significant Feature ◽  
Pressurized Water ◽  
Safety Testing ◽  
Fuel Assemblies ◽  
Enriched Uranium

This article reviews that after a half century of safety testing for the nuclear industry, a key heat-transfer lab is losing its home. Columbia University’s Heat Transfer Research Facility has been the only place to go for key safety testing. Since the days of the Atoms for Peace program during the Eisenhower years, the lab has tested generations of nuclear reactor fuel assemblies. The lab’s clients over the years have included all the designers of pressurized water reactors in the United States and others from much of the world. The tests are primarily concerned with one small, but significant feature of a reactor core. A core contains as many as 3000 fuel assemblies, bundles of long, slender rods containing enriched uranium. Controlled fission among the bundles heats water to begin the series of heat-transfer cycles that send steam to the turbines that will drive generators.

scholarly journals AI-based design of a nuclear reactor core

2021 ◽  
Vol 11 (1) ◽  
Author(s):  
Vladimir Sobes ◽  
Briana Hiscox ◽  
Emilian Popov ◽  
Rick Archibald ◽  
Cory Hauck ◽  
...  
Keyword(s):  
Nuclear Reactor ◽  
Rapid Development ◽  
Reactor Core ◽  
Systems Design ◽  
Axial Direction ◽  
Arbitrary Geometry ◽  
Oak Ridge ◽  
Geometry Design ◽  
Nuclear Systems ◽  
Nuclear Reactor Core

AbstractThe authors developed an artificial intelligence (AI)-based algorithm for the design and optimization of a nuclear reactor core based on a flexible geometry and demonstrated a 3× improvement in the selected performance metric: temperature peaking factor. The rapid development of advanced, and specifically, additive manufacturing (3-D printing) and its introduction into advanced nuclear core design through the Transformational Challenge Reactor program have presented the opportunity to explore the arbitrary geometry design of nuclear-heated structures. The primary challenge is that the arbitrary geometry design space is vast and requires the computational evaluation of many candidate designs, and the multiphysics simulation of nuclear systems is very time-intensive. Therefore, the authors developed a machine learning-based multiphysics emulator and evaluated thousands of candidate geometries on Summit, Oak Ridge National Laboratory’s leadership class supercomputer. The results presented in this work demonstrate temperature distribution smoothing in a nuclear reactor core through the manipulation of the geometry, which is traditionally achieved in light water reactors through variable assembly loading in the axial direction and fuel shuffling during refueling in the radial direction. The conclusions discuss the future implications for nuclear systems design with arbitrary geometry and the potential for AI-based autonomous design algorithms.

scholarly journals CFD Analysis of Downcomer of Nuclear Reactor VVER 440

2016 ◽  
Vol 66 (2) ◽  
pp. 55-62
Author(s):  
Vladimír Kutiš ◽  
Jakub Jakubec ◽  
Juraj Paulech ◽  
Gálik Gálik ◽  
Tibor Sedlár
Keyword(s):  
Nuclear Power ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Coolant Flow ◽  
Orifice Diameter ◽  
Cfd Analysis ◽  
Flow Conditions ◽  
Fuel Assemblies ◽  
Flow Through ◽  
Cfd Analyses

Abstract The paper is focused on CFD analyses of the coolant flow in the nuclear reactor VVER 440. The goal of the analyses is to investigate the influence of the orifice diameter on the mass flow through individual fuel assemblies in the reactor core. The diameter of orifice can be changed during the operation of a nuclear power plant. Considered boundary conditions in the investigated region of the coolant are based on nominal coolant flow conditions in the nuclear reactor VVER 440.

Comparisons of Rotordynamic Characteristics Predictions for Annular Gas Seals Using the Transient Computational Fluid Dynamic Method Based on Different Single-Frequency and Multifrequency Rotor Whirling Models

2015 ◽  
Vol 138 (1) ◽  
Author(s):  
Zhigang Li ◽  
Jun Li ◽  
Zhenping Feng
Keyword(s):  
Computational Fluid Dynamic ◽  
Fluid Dynamic ◽  
Three Dimensional ◽  
Single Frequency ◽  
Solution Technique ◽  
Model Parameters ◽  
Multiple Frequency ◽  
Rotordynamic Coefficients ◽  
Gas Seals ◽  
Cfd Method

The three-dimensional (3D) transient computational fluid dynamic (CFD) method was proposed to predict rotordynamic coefficients for annular gas seals. This transient CFD method uses unsteady Reynolds-Averaged Navier–Stokes (RANS) solution technique and mesh deformation theory, which requires a rotor whirling model as the rotor excitation signal to solve the transient leakage flow field in seal and obtain the transient fluid response forces on the rotor surface. A fully partitioned pocket damper seal (FPDS) was taken as the test object to validate the present numerical method. Comparisons were made between experimental data and rotordynamic coefficient predictions using the three variations of the single-frequency and multiple-frequency rotor whirling models: (1) one-dimensional whirling model, (2) circular orbit whirling model, and (3) elliptical orbit whirling model. The numerical results show that the rotordynamic coefficients predicted by the present CFD method and six different rotor whirling models all agree well with the experiment data, and nearly coincide for all rotor whirling models. The proposed transient CFD method can be used to perform a reasonably accurate prediction of the frequency-dependent rotordynamic coefficients for annular gas seals based on any one of the present six rotor whirling models, as long as ensuring the combination of these whirling model parameters captures the small perturbation theory. The rotor whirling parameters such as whirling orbit, amplitude, and frequency number are important in predicting rotor whirling motion and fluid response forces, but have almost no effect on the computed rotordynamic coefficients. The benefit of the multiple-frequency rotor whirling models is the ability to calculate accurate rotordynamic coefficients of annular gas seals in a wide frequency range with a simulation time on the order of one-tenth the cost of the single-frequency whirling models.

scholarly journals Study of natural uranium fuel for a new reactor design TEPLATOR

2021 ◽  
Vol 253 ◽  
pp. 07012
Author(s):  
Tomas Peltan ◽  
Eva Vilimova ◽  
Radek Skoda
Keyword(s):  
Heavy Water ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
District Heating ◽  
Alternative Fuel ◽  
Natural Uranium ◽  
Thermal Reactor ◽  
Monte Carlo Code ◽  
Fuel Assemblies ◽  
Irradiated Fuel

The TEPLATOR is a new type of nuclear reactor which the main purpose is producing heat for district heating. It is designed as a special thermal reactor with 55 fuel channels for fuel assemblies, which is moderated and cooled by heavy water and operated around atmospheric pressure. The TEPLATOR DEMO is designed for the use of irradiated fuel from PWR or BWR reactors. Using heavy water as the moderator and coolant in this reactor concept allows to use natural uranium as an alternative fuel in case that the irradiated fuel is not available for some reason. This solution is suitable because of the price of natural uranium and the absence of costly fuel enrichment. This article is focused on deeper analyses of alternative suitable fuel for TEPLATOR based on natural uranium and new fuel geometries. This work builds on previous research on alternative fuel material and geometry for the TEPLATOR. It is mainly concerned with the neutronic development of fuel assemblies, the possibility of manufacturing of developed fuel types, and optimization of fuel management and uranium consumption. This article contains predetermined candidates for suitable fuel geometries and new untested fuel geometry types with some new advantages. Finally, optimization of the whole reactor core and number of fuel channels was made in terms of increased safety and higher fuel burn-up. Presented calculations were performed by Monte Carlo code Seprent.

Health Monitoring of Centrifugal Pumps Using Digital Models

2019 ◽  
Vol 141 (9) ◽  
Author(s):  
Amr A. Abdel Fatah ◽  
Mohammed A. Hassan ◽  
Mohamed Lotfy ◽  
Antoine S. Dimitri
Keyword(s):  
Centrifugal Pump ◽  
Fluid Dynamic ◽  
Rapid Development ◽  
Dynamic Performance ◽  
Three Dimensional ◽  
Operating Conditions ◽  
Normal Operation ◽  
Centrifugal Pumps ◽  
Digital Models ◽  
Exciting Forces

The area of predictive maintenance (PM) has received growing research interest in the past few years. Diagnostic capabilities of PM technologies have increased due to advances made in sensor technologies, signal processing algorithms, and the rapid development of computational power and data handling algorithms. Conventional PM programs are mostly built around analyzing sensors' data collected from physical systems. Incorporating simulation data collected from digital models replicating the physical system with sensors' data can lead to more optimization for operation and maintenance. This paper demonstrates the role of using digital models in implementing effective condition monitoring on centrifugal pumps. Two digital models are used to study the dynamic performance of a centrifugal pump experiencing cavitation condition. The first model is a three-dimensional fully turbulent computational fluid dynamic (CFD) model. Based on the pressure distribution obtained from the CFD, a novel analytical pressure pulsation model is developed and used to simulate the exciting forces affecting the pump. The second digital model is a pump casing dynamic model which is used to predict the casing vibration response to exciting forces due to faulty operating conditions. Results obtained from the digital models are validated using an experimental test rig of a small centrifugal pump. Using this concept, a pump faulty operation can be simulated to provide complete understanding of the root cause of the fault. Additionally, digital models can be used to simulate different corrective actions that would restore the normal operation of the pump.

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