FLUENT Analysis of a ROSA Two-Phase Stratification Test

2009 ◽  
Author(s):  
Tatiana Farkas ◽  
Iva´n To´th
Keyword(s):  
Best Practice ◽  
Nuclear Reactor ◽  
Natural Circulation ◽  
Water Injection ◽  
Water Layer ◽  
Two Phase ◽  
Injection Point ◽  
Best Practice Guidelines ◽  
Vof Model ◽  
Nuclear Reactor Safety

One of the OECD ROSA project tests, investigating temperature stratification in the cold legs and the downcomer during ECCS water injection under two-phase natural circulation conditions was analysed with the FLUENT code. The guidance given in the “Best Practice Guidelines for the Use of CFD in Nuclear Reactor Safety Applications” of the OECD GAMA group was followed. The standard k-ε turbulence model of FLUENT was applied along with the VOF (Volume of Fluid) description of the steam/water phases. In order to model the interface of the two phases more closely, transient calculations were performed. Based on a comparison of calculated and measured results, it is found that the standard k-ε turbulence along with the VOF model gave acceptable description of the temperature distribution within the water layer. However, the model under-estimates mixing at the injection point, while it is over-estimated further downstream in the stratified water flow. The condensation model applied in FLUENT strongly under-estimated the subcooling of the steam phase. Insufficient condensation might explain why the downcomer level drops below the cold leg, which was not the case in the test.

Transient Analysis for the Case of Changes in Coolant Inventory for Advanced Heavy Water Reactor

2004 ◽  
Author(s):  
H. G. Lele ◽  
A. Srivastava ◽  
B. Chatterjee ◽  
A. J. Gaikwad ◽  
Rajesh Kumar ◽  
...  
Keyword(s):  
Heavy Water ◽  
Nuclear Reactor ◽  
Natural Circulation ◽  
Reactor Core ◽  
Two Phase ◽  
Decay Heat ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Heavy Water Reactor ◽  
Core Cooling

Safety of nuclear reactor needs to be assessed against different categories of Postulated initiating events. Advanced Heavy Water Reactor is natural circulation light water cooled and heavy water moderated pressure tube type of reactor. Inventory of the system is important parameter in determination of flow characteristics of this natural circulation reactor. In view of this, various events that cause changes in PHT system inventory are analysed in this paper. One of the reason for decrease in coolant inventory is hypothetical Loss of coolant accident (LOCA) This event is of very low probability but important from designing engineered safeguard system of a reactor. Loss of coolant accident in a nuclear reactor can cause voiding of the reactor core due to expulsion of primary coolant from break. In such, a situation the reactor core experiences very low heat removal rate from the nuclear fuel though the decay heat generation continues even after tripping of the reactor. Heat generation in the reactor core is due to various sources such as decay heat, stored heat etc, can lead to heating of fuel elements. However, Emergency core cooling systems of the reactor are actuated and prevent undesirable temperature rise. These events are called design basis events and focus is on adequacy of Emergency Core Cooling System (ECCS) and fuel integrity. The scenarios, phenomena encountered and consequences depend upon size and location of break, system characteristics, and actuation and capability of different protection and engineered safeguard systems of the reactor system. Moreover, this reactor has several passive features to ensure safety of this reactor. which are considered in analyzing these events. Events under category of decrease in coolant inventory includes loss of coolant accidents due to break at different locations of different sizes. Various locations considered in this paper are steam line, inlet header, inlet feeder, ECCS header, downcomer, pressure tube, Isolation condenser inlet header, instrument line break at inlet header and steam drum. The paper also considers scenario emerging due to malfunctions like relief valve stuck open. Causes for events under category of increase in coolant inventory are Increase in Drum level controller set point, Inadvertent valving in of Accumulators and Inadvertent valving in of Gravity driven water pool (GDWP). Last two events are not analysed as they are not possible. The analysis for the above events is complex due to various complex and wide ranges of phenomena involved during different pies under this category. It involves single and two phase natural circulation at different power levels, inventories and pressures, two-phase natural circulation under depleted inventory conditions. Coupled neutronics and thermal hydraulics behaviour, Phenomena under LOCA, phenomena during ECCS injection, direct injection into fuel rod, advanced accumulator injection., vapour pull through and coupled controller and thermal hydraulics. Modelling of these phenomena for each event is discussed in this paper. In this paper summary of analyses for representive event is presented.

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.

Application of CFX-10 to the Investigation of RPV Coolant Mixing in VVER Reactors

2006 ◽  
Author(s):  
Fabio Moretti ◽  
Daniele Melideo ◽  
Fulvio Terzuoli ◽  
Francesco D’Auria
Keyword(s):  
Pressure Vessel ◽  
Nuclear Reactor ◽  
Natural Circulation ◽  
Three Dimensional ◽  
Sensitivity Analyses ◽  
Model Parameters ◽  
Pressurized Water ◽  
Relevant Role ◽  
Nuclear Reactor Safety ◽  
Mixing Problem

Coolant mixing phenomena occurring in the pressure vessel of a nuclear reactor constitute one of the main objectives of investigation by researchers concerned with nuclear reactor safety. For instance, mixing plays a relevant role in reactivity-induced accidents initiated by deboration or boron dilution events, followed by transport of a deborated slug into the vessel of a pressurized water reactor. Another example is constituted by temperature mixing, which may sensitively affect the consequences of a pressurized thermal shock scenario. Predictive analysis of mixing phenomena is strongly improved by the availability of computational tools able to cope with the inherent three-dimensionality of such problem, like system codes with three-dimensional capabilities, and Computational Fluid Dynamics (CFD) codes. The present paper deals with numerical analyses of coolant mixing in the reactor pressure vessel of a VVER-1000 reactor, performed by the ANSYS CFX-10 CFD code. In particular, the “swirl” effect that has been observed to take place in the downcomer of such kind of reactor has been addressed, with the aim of assessing the capability of the codes to predict that effect, and to understand the reasons for its occurrence. Results have been compared against experimental data from V1000CT-2 Benchmark. Moreover, a boron mixing problem has been investigated, in the hypothesis that a deborated slug, transported by natural circulation, enters the vessel. Sensitivity analyses have been conducted on some geometrical features, model parameters and boundary conditions.

scholarly journals Assessment of the MARS Code Using the Two-Phase Natural Circulation Experiments at a Core Catcher Test Facility

2017 ◽  
Vol 2017 ◽  
pp. 1-13
Author(s):  
Dong Hun Lee ◽  
Su Ryong Choi ◽  
Kwang Soon Ha ◽  
Han Young Yoon ◽  
Jae Jun Jeong
Keyword(s):  
Computational Analysis ◽  
Nuclear Reactor ◽  
Natural Circulation ◽  
Three Dimensional ◽  
Heating Surface ◽  
Severe Accident ◽  
Test Facility ◽  
Two Phase ◽  
Core Catcher ◽  
Parametric Effects

A core catcher has been developed to maintain the integrity of nuclear reactor containment from molten corium during a severe accident. It uses a two-phase natural circulation for cooling molten corium. Flow in a typical core catcher is unique because (i) it has an inclined cooling channel with downwards-facing heating surface, of which flow processes are not fully exploited, (ii) it is usually exposed to a low-pressure condition, where phase change causes dramatic changes in the flow, and (iii) the effects of a multidimensional flow are very large in the upper part of the core catcher. These features make computational analysis more difficult. In this study, the MARS code is assessed using the two-phase natural circulation experiments that had been conducted at the CE-PECS facility to verify the cooling performance of a core catcher. The code is a system-scale thermal-hydraulic (TH) code and has a multidimensional TH component. The facility was modeled by using both one- and three-dimensional components. Six experiments at the facility were selected to investigate the parametric effects of heat flux, pressure, and form loss. The results show that MARS can predict the two-phase flow at the facility reasonably well. However, some limitations are obviously revealed.

Development of a Nuclear Reactor Two-Phase Natural Circulation CFD Model

2004 ◽  
Author(s):  
D. A. Botelho ◽  
P. A. B. De Sampaio ◽  
M. L. Moreira ◽  
J. L. H. Faccini
Keyword(s):  
Nuclear Reactor ◽  
Natural Circulation ◽  
Heat Removal ◽  
Two Phase ◽  
One Dimensional ◽  
Temperature Oscillations ◽  
Computational Fluid Dynamics Cfd ◽  
The University ◽  
Engineering Institute ◽  
Method Show

A circuit of natural convection similar to the passive reactor heat removal systems of the AP600 reactor and of the APEX experiment was constructed in the Nuclear Engineering Institute located in Rio de Janeiro. In an attempt to improve usual numeric methods, it was developed a new Computational Fluid Dynamics (CFD) scheme to solve one-dimensional (1D) two-phase natural circulation transport equations. To compare the new scheme with other numeric methods, it was used the results of the two-phase experiment conducted at the University of Sa˜o Paulo (USP). It was compared the frequency and the amplitude of the temperature oscillations at several positions of the spatial domain during a two-phase heating transient. The results of the new method show a better agreement with the experiment than that obtained from the RELAP and the TRAC series of computer programs.

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.

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