Assessment of Post-CHF Heat Transfer Models for the SPACE Code

2009 ◽  
Author(s):  
K. Y. Choi ◽  
B. J. Yun ◽  
H. S. Park ◽  
S. K. Moon ◽  
K. D. Kim
Keyword(s):  
Heat Transfer ◽  
Mode Selection ◽  
Film Boiling ◽  
Model Assessment ◽  
Field Equations ◽  
Pressurized Water ◽  
Water Reactors ◽  
And Performance ◽  
Physical Phenomena ◽  
Transfer Models

The SPACE (Safety and Performance Analysis Code) which is based on a multi-dimensional two-fluid, three-field model is under development for a licensing purpose of pressurized water reactors in Korea. A total of 12 wall-to-fluid heat transfer modes were defined and a heat transfer mode selection logic was developed according to the noncondensable gas quality, the void fraction, the degree of subcooling and the wall temperature. Among the constitutive equations of the SPACE code, post-CHF heat transfer models are considered to have the most significant uncertainties because their physical phenomena are not fully understood yet. Though a variety of models and correlations for the post-CHF heat transfer are available, there is no reliable model to reproduce the heat transfer rate as realistically as possible. Several post-CHF heat transfer models are implemented into the SPACE code; critical heat, transition and film boiling models. The present paper describes the model assessment progress which was done for the transition and film boiling models of the SPACE code. A heat flux partition into the continuous liquid, entrained droplet and vapor fields should be taken into account in order to be in line with physical phenomena as the SPACE code has three-field equations in its hydraulic solver. Existing energy partitioning methods were peer-reviewed in order to determine the best model which can be applicable to the SPACE code. The present work will help to consolidate the developed wall-to-fluid heat transfer package of the SPACE code.

scholarly journals Experimental approach to investigate the dynamics of mixing coolant flow in complex geometry using PIV and PLIF techniques

2015 ◽  
Vol 19 (3) ◽  
pp. 989-1004 ◽  
Author(s):  
Ezddin Hutli ◽  
Valer Gottlasz ◽  
Dániel Tar ◽  
Gyorgy Ezsol ◽  
Gabor Baranyai
Keyword(s):  
Heat Transfer ◽  
Structural Integrity ◽  
Complex Geometry ◽  
Measurement Techniques ◽  
Coolant Flow ◽  
Temperature Fields ◽  
Head Part ◽  
Pressurized Water ◽  
Area Of Interest ◽  
Water Reactors

The aim of this work is to investigate experimentally the increase of mixing phenomenon in a coolant flow in order to improve the heat transfer, the economical operation and the structural integrity of Light Water Reactors-Pressurized Water Reactors (LWRs-PWRs). Thus the parameters related to the heat transfer process in the system will be investigated. Data from a set of experiments, obtained by using high precision measurement techniques, Particle Image Velocimetry and Planar Laser-Induced Fluorescence (PIV and PLIF, respectively) are to improve the basic understanding of turbulent mixing phenomenon and to provide data for CFD code validation. The coolant mixing phenomenon in the head part of a fuel assembly which includes spacer grids has been investigated (the fuel simulator has half-length of a VVER 440 reactor fuel). The two-dimensional velocity vector and temperature fields in the area of interest are obtained by PIV and PLIF technique, respectively. The measurements of the turbulent flow in the regular tube channel around the thermocouple proved that there is rotation and asymmetry in the coolant flow caused by the mixing grid and the geometrical asymmetry of the fuel bundle. Both PIV and PLIF results showed that at the level of the core exit thermocouple the coolant is homogeneous. The discrepancies that could exist between the outlet average temperature of the coolant and the temperature at in-core thermocouple were clarified. Results of the applied techniques showed that both of them can be used as good provider for data base and to validate CFD results.

Development of Simulation Codes for Some Accidents in Pressurized Water Reactors

2008 ◽  
Author(s):  
X. B. Yang ◽  
G. H. Su ◽  
S. Z. Qiu
Keyword(s):  
Heat Transfer ◽  
Physical Property ◽  
Friction Model ◽  
Physical Models ◽  
Pressurized Water Reactors ◽  
Transfer Model ◽  
Feed Water ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactors

An analysis code has been developed for evaluating the transient thermo-hydraulic behaviors of the pressurized water reactor system. A series of mathematical and physical models is considered in this code, such as the point reactor neutron kinetics model, the heat transfer model, the friction model, the thermo-physical property model and so on. All possible flow and heat transfer conditions in some accidents have been considered and their corresponding models are supplied. Gear’s method is adopted for a better solution to the stiff equations. In this paper, some general accidents in the pressurized water reactors have been investigated, including the station blackout accident (SBO), the loss of flow accident (LOFA), the loss of feed water accident (LOFWA) and the reactivity insertion accident (RIA). The calculated results have been verified by the RELAP5/Mod3 and the results are satisfactory.

scholarly journals Physical Phenomena in Nuclear Thermal Hydraulics and Current Status

2021 ◽  
Vol 65 (1) ◽  
pp. 1-11
Author(s):  
F. D’Auria ◽  
N. Aksan ◽  
H. Glaeser
Keyword(s):  
Current Status ◽  
Thermal Hydraulics ◽  
Two Phase ◽  
Energy Agency ◽  
Pressurized Water ◽  
Current State ◽  
Integral Effect ◽  
Water Reactors ◽  
Physical Phenomena ◽  
Effect Test

116 nuclear Thermal-Hydraulic Phenomena T-HP are identified in the present paper, following documents issued during the last three decades by the Committee on the Safety of Nuclear Installations of Nuclear Energy Agency of the Organization for Economic Cooperation and Development (OECD/NEA/CSNI) and by the International Atomic Energy Agency (IAEA). The derived T-HP list includes consideration of experiments performed in Separate Effect Test (SET) and Integral Effect Test (IET) facilities relevant to reactor coolant system and containment of Water Cooled Nuclear Reactors (WCNR). We consider a dozen WCNR types: Pressurized Water Reactors (PWR), Boiling Water Reactors (BWR), Russian reactors (VVER-440, VVER-1000 and RBMK), pressure tube heavy water reactors by Canada (CANDU) and India (PHWR) and so-called ‘advanced’ reactors (e.g. AP-1000 and APR-1400 designed in US and Korea, respectively). We envisage a variety of applications for the T-HP list. Four of the phenomena are helpful to characterize the current state of art in nuclear thermal-hydraulics: Counter Current Flow Limitation (CCFL), Critical Heat Flux (CHF), reflood and Two-Phase Critical Flow (TPCF). Furthermore, the T-HP identification contributes to addressing the scaling issue, performing uncertainty evaluations, developing constitutive equations and ‘special models’ in codes and prioritizing the research.

scholarly journals Zeta Potentials of Magnetite Particles and Alloy 690 Surfaces in Alkaline Solutions

Materials ◽  
2020 ◽  
Vol 13 (18) ◽  
pp. 3999
Author(s):  
Ji-Min Lee ◽  
Dong-Seok Lim ◽  
Soon-Hyeok Jeon ◽  
Do Haeng Hur
Keyword(s):  
Magnetite Nanoparticles ◽  
Ph Value ◽  
Alkaline Solutions ◽  
Empirical Formulas ◽  
Pressurized Water ◽  
Zeta Potentials ◽  
Alloy 690 ◽  
Magnetite Particles ◽  
Water Reactors ◽  
And Performance

Magnetite particles deposited on the secondary side of a steam generator (SG) can degrade the integrity and performance of pressurized water reactors. Therefore, it is necessary to produce the data of fundamental interfacial electrokinetic properties of magnetite particles and SG tube materials. This study investigated the zeta potentials of magnetite nanoparticles and Alloy 690 surfaces, which were dependent on the pH value, pH agent, and the presence of NaCl. The zeta potentials of the magnetite nanoparticles increased in the negative direction as the pH increased, regardless of the pH agent. At the same pH value, the absolute values of the zeta potentials with different pH agents were: ethanolamine < ammonia < morpholine. In the presence of NaCl, the zeta potentials of the particles further increased negatively. The meaning of the measured zeta potentials was discussed in terms of the dispersion stability and the agglomeration of the particles. Based on the relationship between the zeta potentials of the particles and Alloy 690 surfaces, the magnetite deposition on Alloy 690 was also discussed. Furthermore, the empirical formulas for the pH-dependent zeta potentials of magnetite particles in each alkaline solution were suggested.

scholarly journals High to Low pellet cladding gap heat transfer modeling methodology in an uncertainty quantification framework for a PWR Rod Ejection Accident with best estimate coupling

2020 ◽  
Vol 6 ◽  
pp. 56
Author(s):  
Gregory Kyriakos Delipei ◽  
Josselin Garnier ◽  
Jean-Charles Le Pallec ◽  
Benoit Normand
Keyword(s):  
Heat Transfer ◽  
Uncertainty Quantification ◽  
High Fidelity ◽  
Pressurized Water ◽  
Modeling Methodology ◽  
Best Estimate ◽  
Potential Gain ◽  
Uncertainty Analyses ◽  
Water Reactors ◽  
Output Variance

High to Low modeling approaches can alleviate the computationally expensive fuel modeling in nuclear reactor’s transient uncertainty quantification. This is especially the case for Rod Ejection Accident (REA) in Pressurized Water Reactors (PWR) were strong multi-physics interactions occur. In this work, we develop and propose a pellet cladding gap heat transfer (Hgap) High to Low modeling methodology for a PWR REA in an uncertainty quantification framework. The methodology involves the calibration of a simplified Hgap model based on high fidelity simulations with the fuel-thermomechanics code ALCYONE1. The calibrated model is then introduced into the CEA developed CORPUS Best Estimate (BE) multi-physics coupling between APOLLO3® and FLICA4. This creates an Improved Best Estimate (IBE) coupling that is then used for an uncertainty quantification study. The results indicate that with IBE the distance to boiling crisis uncertainty is decreased from 57% to 42%. This is reflected to the decrease of the sensitivity of Hgap. In the BE coupling Hgap was responsible for 50% of the output variance while in IBE it is close to 0. These results show the potential gain of High to Low approaches for Hgap modeling in REA uncertainty analyses.

Sign in / Sign up

Export Citation Format

Share Document