Coupling of the 3D Neutron Kinetic Core Model DYN3D With the CFD Software ANSYS CFX

2014 ◽  
Author(s):  
Alexander Grahn ◽  
Sören Kliem ◽  
Ulrich Rohde
Keyword(s):  
Heat Transfer ◽  
Three Dimensional ◽  
Reactor Core ◽  
General Purpose ◽  
Core Region ◽  
Core Model ◽  
Hydraulic Calculation ◽  
Test Case ◽  
Control Rod ◽  
Ansys Cfx

This article presents the implementation of a coupling between the 3D neutron kinetic core model DYN3D and the commercial, general purpose computational fluid dynamics (CFD) software ANSYS-CFX. In the coupling approach, parts of the thermal hydraulic calculation are transferred to CFX for its better ability to simulate the three-dimensional coolant redistribution in the reactor core region. The calculation of the heat transfer from the fuel into the coolant remains with DYN3D, which incorporates well tested and validated heat transfer models for rod-type fuel elements. On the CFX side, the core region is modelled based on the porous body approach. The implementation of the code coupling is verified by comparing test case results with reference solutions of the DYN3D standalone version. Test cases cover mini and full core geometries, control rod movement and partial overcooling transients.

scholarly journals RAST-K v2—Three-Dimensional Nodal Diffusion Code for Pressurized Water Reactor Core Analysis

Energies ◽  
2020 ◽  
Vol 13 (23) ◽  
pp. 6324
Author(s):  
Jinsu Park ◽  
Jaerim Jang ◽  
Hanjoo Kim ◽  
Jiwon Choe ◽  
Dongmin Yun ◽  
...  
Keyword(s):  
Power Plants ◽  
Single Channel ◽  
Three Dimensional ◽  
Reactor Core ◽  
Reactor Power ◽  
Hydraulic Calculation ◽  
Pressurized Water ◽  
Computational Performance ◽  
Nodal Diffusion ◽  
Calculation Models

The RAST-K v2, a novel nodal diffusion code, was developed at the Ulsan National Institute of Science and Technology (UNIST) for designing the cores of pressurized water reactors (PWR) and performing analyses with high accuracy and computational performance by adopting state-of-the-art calculation models and various engineering features. It is a three-dimensional multi-group nodal diffusion code developed for the steady and transient states using microscopic cross-sections generated by the STREAM code for 37 isotopes. A depletion chain containing 22 actinides and 15 fission products and burnable absorbers was solved using the Chebyshev rational approximation method. A simplified one-dimensional single-channel thermal-hydraulic calculation was performed with various values for the thermal conductivity. Advanced features such as burnup adaptation and CRUD modeling capabilities are implemented for the multi-cycle analysis of commercial reactor power plants. The performance of RAST-K v2 has been validated with the measured data of PWRs operating in Korea. Furthermore, RAST-K v2 has been coupled with a sub-channel code (CTF), fuel performance code (FRAPCON), and water chemistry code for multiphysics analyses. In this paper, the calculation models and engineering features implemented in RAST-K v2 are described, and then the application status of RAST-K v2 is presented.

Monte Carlo Simulation of Radiative Heat Transfer in Rapid Thermal Processing (RTP) Systems

1994 ◽  
Vol 342 ◽  
Author(s):  
J. Vernon Cole ◽  
Karson L. Knutson ◽  
Klavs F. Jensen
Keyword(s):  
Heat Transfer ◽  
Monte Carlo ◽  
Thermal Processing ◽  
Measurement Techniques ◽  
Radiation Heat Transfer ◽  
Rapid Thermal Processing ◽  
Three Dimensional ◽  
General Purpose ◽  
Radiation Heat ◽  
Participating Media

ABSTRACTWe present a general purpose Monte Carlo method for the simulation of radiation heat transfer in rapid thermal processing (RTP) chambers. Three-dimensional mesh generation software is used to discretize the surfaces within the system, allowing the simulation of realistic chamber and reflector designs. An adaptive subdivision of the chamber geometry reduces the number of raysurface intersections which must be computed. The method models internal reflection, absorption, and transmission within participating media, and includes wavelength, temperature, and material dependent optical properties. Radiation heat transfer simulations are used to examine a reflector assembly, and to test the assumptions of optical wafer temperature measurement techniques.

Study on the Laminar Mixed Convective Heat Transfer in Three-Dimensional Channel With a Thermal Source

1991 ◽  
Vol 113 (1) ◽  
pp. 40-49 ◽  
Author(s):  
Heiu-Jou Shaw ◽  
Wen-Lih Chen ◽  
Cha’o-Kuang Chen
Keyword(s):  
Heat Transfer ◽  
Convective Heat Transfer ◽  
Convective Heat ◽  
Three Dimensional ◽  
General Purpose ◽  
Heating Element ◽  
Thermal Source ◽  
Simpler Algorithm ◽  
Alternating Direction ◽  
General Purpose Computer

In this paper, the mixed convective heat transfer phenomena in a three-dimensional channel with one heating-element is studied. A general-purpose computer program is developed to analyze the flow field and temperature distribution in the channel. In order to ensure the accuracy of result, numerical computation is performed by using the alternating direction implicit (A.D.I.) method based on the SIMPLER algorithm. This paper deals with fundamental heat transfer phenomena in “L.S.I.” package, which is used extensively in microelectronic equipment. The influences of Reynolds number and Grashof number on the Nusselt number of the heating-element are discussed. In order to make it easier for readers to understand the phenomena studied in this paper, three-dimensional streaklines and three-dimensional isothermal surfaces are presented.

scholarly journals Three-dimensional time-dependent neutron diffusion simulation using average current nodal expansion method

2020 ◽  
Vol 35 (3) ◽  
pp. 189-200
Author(s):  
Kambiz Valavi ◽  
Ali Pazirandeh ◽  
Gholamreza Jahanfarnia
Keyword(s):  
Three Dimensional ◽  
Reactor Core ◽  
Expansion Method ◽  
Steady State Solution ◽  
Time Dependent ◽  
Test Case ◽  
Neutron Diffusion ◽  
Average Current ◽  
Coarse Meshes ◽  
Nodal Expansion Method

In this work, the average current nodal expansion method was developed for the time-dependent neutronic simulation of transients in a nuclear reactor's core. For this purpose, an adopted iterative algorithm was proposed for solving the 3-D time-dependent neutron diffusion equation. In the average current nodal expansion method, the domain of the reactor core can be modeled by coarse meshes for neutronic calculation associated with reasonable precision of results. The discretization of time differential terms in the time-dependent equations was fulfilled, according to the implicit scheme. The proposed strategy was implemented in some kinetic problems including an infinite slab reactor, TWIGL 2-D seed-blanket reactor, and 3-D LMW LWR. At first, the steady-state solution was carried out for each test case, and then, the dynamic neutronic calculation was performed during the time for a specified transient scenario. Obtained results of static and dynamic solutions were verified in comparison with well-known references. Results can indicate the ability of the developed calculator to simulate transients in a nuclear reactor's core.

Comparative Study of Optimization in Pultrusion with Pre-Heating and Die-Cooler Temperature for Improved Cure

2016 ◽  
Vol 879 ◽  
pp. 402-407
Author(s):  
Rita de Cassia Dias Costa ◽  
Lizandro de Sousa Santos ◽  
Ralf Schledjewski
Keyword(s):  
Recent Literature ◽  
Three Dimensional ◽  
General Purpose ◽  
Heat Transfer Analysis ◽  
Process Time ◽  
Conductive Heat ◽  
Mass Balances ◽  
Degree Of Cure ◽  
Ansys Cfx ◽  
The Mean

Pultrusion is a composite manufacturing technique for processing continuous composite profiles with a constant cross section. In such system, energy and mass balances are used to model the thermal and kinetic behavior of the material during processing. This work aims to compare the results obtained in the recent literature, regarding thermal optimization of pultrusion. In the present analysis, an alternative thermal configuration has been suggested, with the objective of maximizing the mean degree of cure. A general-purpose FE software, ANSYS-CFX®, has been used to perform a three-dimensional (3D) conductive heat transfer analysis. Several case studies were conducted where the degree of cure was analyzed for varying heating scenarios. Results have shown that it is possible to get a higher cure in less process time if the die is isolated from the environment.

Development of an Efficient Three-Dimensional Reactor Core Model for Simulator Applications

1992 ◽  
Vol 97 (3) ◽  
pp. 352-361 ◽  
Author(s):  
Mankit Ray Yeung ◽  
Guo Bing Jiang
Keyword(s):  
Three Dimensional ◽  
Reactor Core ◽  
Core Model

Validation of TRACG05 for Application to BWR AOO

2012 ◽  
Author(s):  
Asuka Matsui ◽  
Masashi Tamitani ◽  
Yoshiro Kudo ◽  
Sho Takano ◽  
Tatsuya Iwamoto ◽  
...  
Keyword(s):  
Experimental Data ◽  
Three Dimensional ◽  
Reactor Core ◽  
Expansion Method ◽  
Thermal Hydraulics ◽  
Control Rod ◽  
Neutron Kinetics ◽  
Safety Issues ◽  
Best Estimate ◽  
Assessment Matrix

TRACG code, coupling a three-dimensional neutron kinetics model for the reactor core with thermal-hydraulics based on two-fluid conservation equations, is a best-estimate (BE) code for BWRs to realistically simulate their transient and accidental behaviors. TRACG05 is the latest version and was originally developed to analyze Reactivity Initiated Accident (RIA). TRACG05 incorporates the same neutronics model of the latest core simulator with a three-group analytic-polynomial nodal expansion method. In addition to application to RIA safety analyses, TRACG05 has been planned to apply to safety analyses for Anticipated Operational Occurrences (AOOs) in BWRs by using a Best Estimate Plus Uncertainty (BEPU) methodology. To apply BEPU with TRACG05 to BWR AOOs, validations must be performed to evaluate the uncertainties of models relevant to important phenomena by comparing with appropriate test results for BWR AOOs. At first, a PIRT (Phenomena Identification and Ranking Table) was developed for each event scenario in AOOs to identify relevant physical processes and to determine their relative importance. According to the PIRT, an assessment matrix was established for separate effects tests (SETs), component effects tests (CETs), integral effects tests (IETs), and integral BWR plant start-up tests. The assessment matrix related the important phenomena to the test database, which was confirmed that all the important phenomena were covered by all tests specified in the matrix. According to the assessment matrix, comparison analyses have been specified to perform systematic and comprehensive validations of TRACG05 applicability to AOOs. The comparison analyses were done as the integrated code system with the up-stream reactor core design codes, therefore higher quality was enabled to evaluate the safety parameters. As the result, the uncertainties of important models in TRACG05 were determined so as to enable BEPU approaches for AOO safety issues. Here, as a SET, comparisons between TRACG05 and experimental data of void fraction in a bundle simulating an actual fuel bundle, which is one of the most important models in the application of TRACG05 to AOO analyses are shown. In addition, as pressurization event in AOOs, comparisons between TRACG05 and experimental data of Peach Bottom 2 Turbine Trip Test, which is one of integral tests for a BWR plant, are shown. This is the only test showing large neutron flux increase and strong coupling of neutron kinetics and thermal-hydraulics in the core due to void and Doppler feedbacks. Furthermore, a sensitivity analysis regarding a delay time of control rod (CR) insertion initiation which was the most sensitive uncertainty to the results is also shown.

Sign in / Sign up

Export Citation Format

Share Document