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.

Computational Analysis of Downcomer Boiling Phenomena Using a Component Thermal Hydraulic Analysis Code, CUPID

2010 ◽  
Vol 133 (5) ◽  
Author(s):  
Hyoung Kyu Cho ◽  
Byong Jo Yun ◽  
Ik Kyu Park ◽  
Jae Jun Jeong
Keyword(s):  
Computational Analysis ◽  
Nuclear Reactor ◽  
Three Dimensional ◽  
Reactor Vessel ◽  
Physical Models ◽  
Low Flow ◽  
Two Phase ◽  
Rate Condition ◽  
Flow Models ◽  
Loss Of Coolant Accident

For the analysis of transient two-phase flows in nuclear reactor components such as a reactor vessel, a steam generator, and a containment, KAERI has developed a three-dimensional thermal hydraulic code, CUPID. It adopts a three-dimensional, transient, two-phase and three-field model and includes various physical models and correlations of the interfacial mass, momentum, and energy transfer for the closure. In the present paper, the CUPID code and its two-phase flow models were assessed against the downcomer boiling experiment, which was performed to simulate the downcomer boiling phenomena. They may happen in the downcomer of a nuclear reactor vessel during the reflood phase of a postulated loss of coolant accident. The stored energy release from the reactor vessel to the liquid inside the downcomer causes the boiling on the wall, and it can reduce the hydraulic head of the accumulated water, which is the driving force of water reflooding to the core. The computational analysis using the CUPID code showed that it can appropriately predict the multidimensional boiling phenomena under a low pressure and low flow rate condition with modification of the bubble size model.

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.

scholarly journals CFD Code Validation against Stratified Air-Water Flow Experimental Data

2008 ◽  
Vol 2008 ◽  
pp. 1-7 ◽  
Author(s):  
F. Terzuoli ◽  
M. C. Galassi ◽  
D. Mazzini ◽  
F. D'Auria
Keyword(s):  
Experimental Data ◽  
Numerical Simulations ◽  
Water Flow ◽  
Nuclear Reactor ◽  
Cold Water ◽  
Three Dimensional ◽  
Free Surface Flow ◽  
Surface Flow ◽  
Two Phase ◽  
Code Validation

Pressurized thermal shock (PTS) modelling has been identified as one of the most important industrial needs related to nuclear reactor safety. A severe PTS scenario limiting the reactor pressure vessel (RPV) lifetime is the cold water emergency core cooling (ECC) injection into the cold leg during a loss of coolant accident (LOCA). Since it represents a big challenge for numerical simulations, this scenario was selected within the European Platform for Nuclear Reactor Simulations (NURESIM) Integrated Project as a reference two-phase problem for computational fluid dynamics (CFDs) code validation. This paper presents a CFD analysis of a stratified air-water flow experimental investigation performed at the Institut de Mécanique des Fluides de Toulouse in 1985, which shares some common physical features with the ECC injection in PWR cold leg. Numerical simulations have been carried out with two commercial codes (Fluent and Ansys CFX), and a research code (NEPTUNE CFD). The aim of this work, carried out at the University of Pisa within the NURESIM IP, is to validate the free surface flow model implemented in the codes against experimental data, and to perform code-to-code benchmarking. Obtained results suggest the relevance of three-dimensional effects and stress the importance of a suitable interface drag modelling.

Heat Removal Capability of Core-Catcher in Consideration of Transformation of Cooling Channel

2014 ◽  
Author(s):  
Tomohisa Kurita ◽  
Mitsuo Komuro ◽  
Ryo Suzuki ◽  
Masato Yamada ◽  
Mika Tahara ◽  
...  
Keyword(s):  
Flow Direction ◽  
Natural Circulation ◽  
Heat Removal ◽  
Nucleate Boiling ◽  
Severe Accident ◽  
Flow Section ◽  
Cooling Channel ◽  
Flow Area ◽  
The Core ◽  
Core Catcher

It is necessary to stabilize high temperature molten core in a severe accident for long time without electrical power. The core-catcher is to be installed at the bottom of the lower drywell in order to settle the molten core flowing down from a reactor vessel. Toshiba’s core-catcher system consists of a round basin made up of inclined cooling channels to get natural circulation of the flooding water. So it can cover all pedestal floor and can work in passive manner. We have been confirming an applicability of the core-catcher to actual plants. We have conducted full scaled tests with a unique cooling channel which has inclined rectangular flow section and changing the section area along flow direction in several conditions to evaluate the influence of the parameters on the natural circulation and heat removal capability. The test results showed good heat removal performance with nucleate boiling. However, we should consider a transformation of the cooling channel, for example, by the falling corium. So we calculate the assumed transformation of the cooling channel and conduct natural circulation tests with obstruction in the cooling channel. We confirm that natural circulation flow is stably continues and the cooling channel can remove prescribed heat, even if a flow area have got narrow locally.

Particle Size Characteristics of Molten Corium Quenched in Water

2008 ◽  
Author(s):  
B. T. Min ◽  
H. D. Kim ◽  
J. H. Kim ◽  
S. W. Hong ◽  
I. K. Park
Keyword(s):  
Particle Size ◽  
Size Distribution ◽  
Steam Explosion ◽  
Eutectic Composition ◽  
Nuclear Reactor ◽  
Cold Water ◽  
Severe Accident ◽  
Test Facility ◽  
Water Steam ◽  
Explosive Composition

During a hypothetical severe accident in a nuclear reactor, a steam explosion might occur when molten corium interacts with water. The strength of a steam explosion affects the integrity of the containment of a nuclear reactor and is highly dependant on the characteristics of the melt-water-steam mixture. Since a break-up and fragmentation process during a pre-mixing are important mechanisms for a steam explosion behavior and affect the debris size distribution, the particle size characteristics of quenched corium have been investigated. For several years, series of experiments have been performed using prototypical corium in the TROI test facility with a high frequency induction heating using cold crucible technology. The molten corium was discharged into the cold water and the quenched debris particles were collected, sieved and examined for the effect of a size distribution on a steam explosion. The small corium droplets do not seem to contribute to a steam explosion owing to solidification at an early stage before the explosion but the large droplets contribute to it owing to their liquid state. It was also shown that single oxides and binary oxides with an eutectic composition (UO2/ZrO2 = 70/30 at weight percentage) led to steam explosions, but a binary oxide with a non-eutectic one did not. The mass mean diameters of the debris of the steam explosive composition was less than that of the non-steam explosive composition. Zirconia was the most energetic steam-explosive material in these tests, and an eutectic composition of corium also lead to a steam explosion, but a non-eutectic composition corium hardly led to a steam explosion. The particle sizes of the molten corium participating in a steam explosion were shown to be mainly 3–6 mm depending on the material and composition.

Experimental and Numerical Investigation on a Two-Phase Natural Circulation Test Facility

2012 ◽  
Vol 33 (9) ◽  
pp. 775-785 ◽  
Author(s):  
Pathayapurayil Pradeep Kumar ◽  
Amod Khardekar ◽  
Kannan N. Iyer
Keyword(s):  
Numerical Investigation ◽  
Natural Circulation ◽  
Test Facility ◽  
Two Phase

Experimental Investigation on Critical Heat Flux From Downward-Facing Flat Plate for Different Orientation Angles

2018 ◽  
Author(s):  
K. H. Deng ◽  
Y. Zhang ◽  
C. L. Wang ◽  
Y. P. Zhang ◽  
W. X. Tian ◽  
...  
Keyword(s):  
Heat Flux ◽  
Critical Heat Flux ◽  
Pressure Vessel ◽  
Nuclear Reactor ◽  
Heating Surface ◽  
Severe Accident ◽  
Width Ratio ◽  
Lower Head ◽  
Length Width ◽  
Core Materials

After the severe accident inside a nuclear reactor, the IVR (In-vessel retention) management strategy is an effective way to keep the integrity of pressure vessel and reduce risk of radioactive leakage by holding the damaged core materials through External Reactor Vessel Cooling (ERVS). The damaged core materials aggregate in the lower head of pressure vessel and releasing heat to the lower head. Therefore, it is very important to remove heat timely to keep the integrity of pressure vessel by ERVS. The shape of lower head is hemispherical and the local Critical Heat Flux (CHF) of different parts changed with latitude. In this paper, influence of orientation angles, area and length-width ratio on CHF of plate heating surface for saturated pool boiling is investigate experimentally. The results show that CHF increases with increasing orientation angles and decreasing area, meanwhile, length-width ratio has a significantly effect on CHF.

1205 An Experimental Study of Two-Phase Natural Circulation System Cooling a Pin Type Finned Heating Surface

2005 ◽  
Vol 2005.80 (0) ◽  
pp. _12-9_-_12-10_
Author(s):  
Mitsunari MASAKA ◽  
Hiroaki KUTSUNA ◽  
Yoshiki NOGUCHI ◽  
Yoichi SHIOMI ◽  
Shigeyasu NAKANISHI
Keyword(s):  
Experimental Study ◽  
Natural Circulation ◽  
Heating Surface ◽  
Circulation System ◽  
Two Phase
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