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

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.

Criticality Calculations and Basic Sensitivity/Uncertainty Investigation of LR-0 Benchmark Core

2019 ◽  
Vol 5 (3) ◽  
Author(s):  
Tomáš Czakoj ◽  
Evžen Losa
Keyword(s):  
Lithium Fluoride ◽  
Three Dimensional ◽  
Reactor Core ◽  
Nuclear Data ◽  
Monte Carlo Code ◽  
Multiplication Factor ◽  
Reference Case ◽  
Energy Version ◽  
Data Libraries ◽  
Nuclear Data Libraries

Three-dimensional Monte Carlo code KENO-VI of SCALE-6.2.2 code system was applied for criticality calculation of the LR-0 reactor core. A central module placed in the center of the core was filled by graphite, lithium fluoride-beryllium fluoride (FLIBE), and lithium fluoride-sodium fluoride (FLINA) compounds. The multiplication factor was obtained for all cases using both ENDF/B-VII.0 and ENDF/B-VII.1 nuclear data libraries. Obtained results were compared with benchmark calculations in the MCNP6 using ENDF/B-VII.0 library. The results of KENO-VI calculations are found to be in good agreement with results obtained by the MCNP6. The discrepancies are typically within tens of pcm excluding the case with the FLINA filling. Sensitivities and uncertainties of the reference case with no filling were determined by a continuos-energy version of the TSUNAMI sequence of SCALE-6.2.2. The obtained uncertainty in multiplication factor due to the uncertainties in nuclear data is about 650 pcm with ENDF/B-VII.1.

CFD Investigation of Thermal-Hydraulic Behaviors in Full Reactor Core for Sodium-Cooled Fast Reactor

2018 ◽  
Author(s):  
Jing Chen ◽  
Dalin Zhang ◽  
Suizheng Qiu ◽  
Kui Zhang ◽  
Mingjun Wang ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Power Distribution ◽  
Three Dimensional ◽  
Reactor Core ◽  
Heat Removal ◽  
The Core ◽  
Computational Fluid Dynamics Cfd ◽  
Heat Removal System ◽  
China Experimental Fast Reactor ◽  
Full Core

As the first developmental step of the sodium-cooled fast reactor (SFR) in China, the pool-type China Experimental Fast Reactor (CEFR) is equipped with the openings and inter-wrapper space in the core, which act as an important part of the decay heat removal system. The accurate prediction of coolant flow in the reactor core calls for complete three-dimensional calculations. In the present study, an investigation of thermal-hydraulic behaviors in a 180° full core model similar to that of CEFR was carried out using commercial Computational Fluid Dynamics (CFD) software. The actual geometries of the peripheral core baffle, fluid channels and narrow inter-wrapper gap were built up, and numerous subassemblies (SAs) were modeled as the porous medium with appropriate resistance and radial power distribution. First, the three-dimensional flow and temperature distributions in the full core under normal operating condition are obtained and quantitatively analyzed. And then the effect of inter-wrapper flow (IWF) on heat transfer performance is evaluated. In addition, the detailed flow path and direction in local inter-wrapper space including the internal and outlet regions are captured. This work can provide some valuable understanding of the core thermal-hydraulic phenomena for the research and design of SFRs.

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.

Real-Time Advanced Nuclear Reactor Core Model

1990 ◽  
pp. 159-169
Author(s):  
Jean Koclas ◽  
F. Friedman ◽  
C. Paquette ◽  
P. Vivier
Keyword(s):  
Real Time ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Core Model ◽  
Nuclear Reactor Core

Parametric Analysis of a Turbine Trip Event in a BWR Using a 3D Nodal Code

2006 ◽  
Author(s):  
A. Gorzel
Keyword(s):  
Three Dimensional ◽  
Reactor Core ◽  
Reactor Power ◽  
Pressure Increase ◽  
Thermal Hydraulics ◽  
Short Term ◽  
Speed Reduction ◽  
Pump Speed ◽  
Transient Behaviour ◽  
Limit Value

Two essential thermal hydraulics safety criteria concerning the reactor core are that even during operational transients there is no fuel melting and impermissible cladding temperatures are avoided. A common concept for boiling water reactors is to establish a minimum critical power ratio (MCPR) for steady state operation. For this MCPR it is shown that only a very small number of fuel rods suffers a short-term dryout during the transient. It is known from experience that the limiting transient for the determination of the MCPR is the turbine trip with blocked bypass system. This fast transient was simulated for a German BWR by use of the three-dimensional reactor analysis transient code SIMULATE-3K. The transient behaviour of the hot channels was used as input for the dryout calculation with the transient thermal hydraulics code FRANCESCA. By this way the maximum reduction of the CPR during the transient could be calculated. The fast increase in reactor power due to the pressure increase and to an increased core inlet flow is limited mainly by the Doppler effect, but automatically triggered operational measures also can contribute to the mitigation of the turbine trip. One very important method is the short-term fast reduction of the recirculation pump speed which is initiated e. g. by a pressure increase in front of the turbine. The large impacts of the starting time and of the rate of the pump speed reduction on the power progression and hence on the deterioration of CPR is presented. Another important procedure to limit the effects of the transient is the fast shutdown of the reactor that is caused when the reactor power reaches the limit value. It is shown that the SCRAM is not fast enough to reduce the first power maximum, but is able to prevent the appearance of a second — much smaller — maximum that would occur around one second after the first one in the absence of a SCRAM.

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