scholarly journals Application of CFD Codes in Nuclear Reactor Safety Analysis

2010 ◽  
Vol 2010 ◽  
pp. 1-8 ◽  
Author(s):  
T. Höhne ◽  
E. Krepper ◽  
U. Rohde
Keyword(s):  
Nuclear Reactor ◽  
Structural Integrity ◽  
Safety Analysis ◽  
Insulation Material ◽  
Reactor Safety ◽  
Two Phase ◽  
Rod Bundle ◽  
Water Reactors ◽  
Nuclear Reactor Safety ◽  
The University

Computational Fluid Dynamics (CFD) is increasingly being used in nuclear reactor safety (NRS) analyses as a tool that enables safety relevant phenomena occurring in the reactor coolant system to be described in more detail. Numerical investigations on single phase coolant mixing in Pressurised Water Reactors (PWR) have been performed at the FZD for almost a decade. The work is aimed at describing the mixing phenomena relevant for both safety analysis, particularly in steam line break and boron dilution scenarios, and mixing phenomena of interest for economical operation and the structural integrity. For the experimental investigation of horizontal two phase flows, different non pressurized channels and the TOPFLOW Hot Leg model in a pressure chamber was build and simulated with ANSYS CFX. In a common project between the University of Applied Sciences Zittau/Görlitz and FZD the behaviour of insulation material released by a LOCA released into the containment and might compromise the long term emergency cooling systems is investigated. Moreover, the actual capability of CFD is shown to contribute to fuel rod bundle design with a good CHF performance.

Application of a new drag coefficient model at CFD-simulations on free surface flows relevant for the nuclear reactor safety analysis

2012 ◽  
Vol 39 (1) ◽  
pp. 70-82 ◽  
Author(s):  
Deendarlianto ◽  
Thomas Höhne ◽  
Pavel Apanasevich ◽  
Dirk Lucas ◽  
Christophe Vallée ◽  
...  
Keyword(s):  
Free Surface ◽  
Drag Coefficient ◽  
Nuclear Reactor ◽  
Safety Analysis ◽  
Free Surface Flows ◽  
Reactor Safety ◽  
Cfd Simulations ◽  
Surface Flows ◽  
Nuclear Reactor Safety

scholarly journals Exploratory Nuclear Reactor Safety Analysis and Visualization via Integrated Topological and Geometric Techniques

2013 ◽  
Author(s):  
Dan Maljovec ◽  
Bei Wang ◽  
Valerio Pascucci ◽  
Peer-Timo Bremer ◽  
Diego Mandelli ◽  
...  
Keyword(s):  
Nuclear Reactor ◽  
Safety Analysis ◽  
Reactor Safety ◽  
Nuclear Reactor Safety ◽  
Geometric Techniques

scholarly journals Processes and Procedures for Application of CFD to Nuclear Reactor Safety Analysis

2006 ◽  
Author(s):  
Richard W. Johnson ◽  
Richard R. Schultz ◽  
Patrick J. Roache ◽  
Ismail B. Celik ◽  
William D. Pointer ◽  
...  
Keyword(s):  
Nuclear Reactor ◽  
Safety Analysis ◽  
Reactor Safety ◽  
Nuclear Reactor Safety

scholarly journals The Best Estimate Plus Uncertainty approach in nuclear reactor safety and licensing: Brief history and the elements after licensing

2019 ◽  
Vol 34 (3) ◽  
pp. 299-312
Author(s):  
Francesco D’Auria ◽  
Giorgio Galassi
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Nuclear Reactors ◽  
Thermal Hydraulics ◽  
Reactor Safety ◽  
Two Phase ◽  
Best Estimate ◽  
Nuclear Reactor Safety ◽  
The Impact

The best estimate plus uncertainty is, at the same time, an approach, a procedure and a frame- work in nuclear thermal-hydraulics and nuclear reactor safety and licensing. The motivation at the basis of the best estimate plus uncertainty is the lack of knowledge in the areas of single and, mainly, two-phase transient thermal-hydraulics. In other terms and introducing some simplifications, the insufficient knowledge of turbulence imposes the design of roadmaps for the application of imperfect (thermal-hydraulic) models to the evaluation of design features and of safety for complex technological installations or systems like the nuclear power plants and, more specifically, the water cooled nuclear reactors. Furthermore, the legal counterpart of nuclear reactor safety, or the licensing, is concerned: therefore the best estimate plus uncertainty must account for rules and regulations derived from the fundamental radioprotection principle which imposes the minimization of the impact of radiations upon humans and the environment under any circumstance. In the present paper, the key elements of the approach are identified and characterized. These shall be seen as the support for a consistent application of thermal-hydraulics to the design and safety of water-cooled nuclear reactors.

Nonhyperbolicity of Conservation Equations of RELAP5 Two-Fluid Model in Nuclear Reactor Safety Results (Investigation and Eigenvalue Analysis)

2020 ◽  
Vol 7 (1) ◽  
Author(s):  
S. P. Saraswat ◽  
P. Munshi ◽  
C. Allison
Keyword(s):  
Nuclear Reactor ◽  
Safety Analysis ◽  
Operating Conditions ◽  
Reactor Safety ◽  
Governing Equations ◽  
Conservation Equations ◽  
Water Reactor ◽  
Code Model ◽  
Nuclear Reactor Safety ◽  
Two Fluid

Abstract The RELAP5 code simulates the thermal-hydraulic characteristics of nuclear reactors by the use of a two-fluid one-dimensional, nonequilibrium, nonhomogeneous two-phase flow model. This model consists of six governing equations to describe the mass, energy, and momentum of the two fluids. The scope of this work comprises the study of the mathematical nature of the code model and to predict the accuracy of the model in the nuclear reactor safety analysis. The method of characteristics (MOC) is applied to check the nonhyperbolic nature of conservation equations for all normal and accident conditions of light water reactors (LWRs). The analysis also gives information about the soundness of the model and to identify the regions where the solutions obtained from it will be numerically convergent. The characteristics of equations of nonhyperbolic nature are complex. It implies that results thus obtained (by finite difference method) have to be interpreted very carefully in view of the sensitive nature of reactor safety analysis. The present analysis shows that governing equations of the code exhibit complex characteristics for some operating conditions thus implying nonhyperbolicity under those conditions. Results are less accurate under such conditions, so sensitivity analysis plays an important role. The sensitivity of closure relationship on the conservation equation's stability is also checked. The analysis is performed in matlab environment for three different systems, (a) pressurized water reactor (PWR), boiling water reactor (BWR), and (c) natural circulation reactor or advance heavy water reactor (AHWR). These results can also be extended to other thermal-hydraulic systems. The different values of the coefficient of closure relationship are taken for different flow regimes. It is observed that the coefficient of virtual mass (for momentum equation) has a significant effect on the hyperbolicity of the system. It is recommended that further development of the RELAP5 model be performed to identify changes that would reduce the region of complex characteristics. The importance of MOC (in nuclear reactor thermal-hydraulic safety analysis) is evident here. In addition, a detailed analysis for operating pressures range of 0.1–22.5 MPa is also performed to find out the nonhyperbolic regions of code model and realistic data of the different type of reactors is used as input of the code. It is also observed here that RELAP5 results are less accurate when system pressure exceeds 19.5 MPa.

Some Nuclear Reactor Safety Related Aspects of Plunging Jets

2010 ◽  
Author(s):  
Eckhard Krepper ◽  
Frank-Peter Weiss ◽  
So¨ren Alt ◽  
Alexander Kratzsch ◽  
Stefan Renger ◽  
...  
Keyword(s):  
Nuclear Reactor ◽  
Reactor Safety ◽  
Joint Research ◽  
Cfd Models ◽  
Safety Research ◽  
Joint Research Project ◽  
Plunging Jets ◽  
Nuclear Reactor Safety ◽  
Entrained Air ◽  
The University

Plunging jets play an important role in nuclear reactor safety research. In the present paper the case of the strainer clogging issue is considered. Entrained air caused by a plunging jet has an influence of the liquid flow field and on the fibre transport in the sump. In the paper the amount of entrained air is given as an inlet boundary condition according to correlations in the literature and confirmed by own experiments. The influence of entrained air on the fibre deposition pattern at the bottom of a tank and on the mixing procedure for the case of temperature differences between jet and tank water are investigated by CFD calculations and compared to experiments. The presented work is part of a joint research project performed in cooperation between the University of Applied Science Zittau/Go¨rlitz and Forschungszentrum Dresden-Rossendorf. The project deals with the experimental investigation of particle transport phenomena in coolant flow in Zittau and the development of CFD models for its simulation in Rossendorf (Krepper et al. 2008).

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