Void Fraction Predictions in Rod Bundles at Low-Pressure Low-Flow Conditions Based on Cunningham-Yeh Model

2004 ◽  
Author(s):  
Cesare Frepoli ◽  
Katsuhiro Ohkawa
Keyword(s):  
Void Fraction ◽  
Pressure Range ◽  
Reactor Core ◽  
Atmospheric Condition ◽  
Original Data ◽  
Low Pressure ◽  
Full Scale ◽  
Low Flow ◽  
Rod Bundles ◽  
The Core

Many experiments have been conducted in the past with full-scale rod bundles to develop void fraction correlations or interfacial drag model which can be used to predict the mixture level in a reactor core following a postulated Loss of Coolant Accident (LOCA). The Cunningham and Yeh correlation was originally developed and validated with boil-off data obtained in the 100 to 400 psia pressure range. Subsequently the validity of the correlation was successfully assessed against several other experiments. However most of the data concentrated in the intermediate to high pressure range (from 100 to 2200 psia). More recently, the development of advanced passive plant such as the AP1000, put more emphasis in the level swell behavior in the near-atmospheric pressure conditions. Following a postulated SBLOCA event for the AP1000 design, the reactor vessel is automatically depressurized to a near atmospheric condition and in the long term the core decay heat is removed by gravity driven injection flow while boiling is occurring in the core. In this paper the Cunningham-Yeh correlation was assessed against data beyond its original data base. Cunningham-Yeh model predictions were compared to several low-flow, low-pressure full-scale rod bundle experiments. Results show that the correlation performs relatively well against low pressure test data. However the Cunningham-Yeh model has the tendency to underpredict the void fraction and therefore to provide conservative results of level swell for plant safety analysis.

LBLOCA Analysis of CPR1000 NPP With Advanced Accumulator

2013 ◽  
Author(s):  
Hongwei Hu ◽  
Jianqiang Shan ◽  
Junli Gou ◽  
Bo Zhang ◽  
Haitao Wang ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Reactor Core ◽  
Low Pressure ◽  
Point Of View ◽  
Primary Coolant ◽  
Pressure Drops ◽  
The Core ◽  
Injection Pump ◽  
Model Combining ◽  
Large Flow

Large break LOCA (LBLOCA) is one of the limit design basic accidents in nuclear power plant. The large flow water in the advanced accumulator is injected into primary loop in early short time. When the vessel pressure drops and reactor core is re-flooded, the advanced accumulator provides a small injection flow to keep the reactor core in flooded condition. Thus, the startup grace time of the low pressure safety injection pump is extended, and the core still stays in a long-term cooling state. By deducing the original accumulator model in RELAP5 accident analysis code, a new model combining the advanced and the traditional accumulator is obtained and coupled into RELAP5/ MOD 3.3. Simulation results show that there is a large flow in the advanced accumulator at the initial stage. When the accumulator water level is lower than the stand pipe, a vortex appears in the damper, resulting in a large pressure drop and small flow. The phenomenon meets the demand of the advanced accumulator design and the simulation of the advanced accumulator is accomplished successfully. Based on this, the primary coolant loop cold leg double-ended guillotine break LBLOCA in CPR1000 is analyzed with the modified RELAP5 code. When the double ended cold leg guillotine accident with 200s delayed startup of the low pressure safety injection occurs, maximum cladding temperature in the core with traditional accumulator is 1860K which seriously exceeded the safety temperature (1477K)[1] prescribed limits while the maximum cladding temperature with advanced accumulator has the security temperature-1277K. From this point of view, adopting passive advanced accumulator can strive a longer grace time for LPSI. Thus the reliability, security and economy of reactor system were improved.

Dynamic Viscosity of Compressed Water to 10 Kilobar and Steam to 1500°C

1969 ◽  
Vol 11 (2) ◽  
pp. 189-205 ◽  
Author(s):  
E. A. Bruges ◽  
M. R. Gibson
Keyword(s):  
Gaseous Phase ◽  
Dynamic Viscosity ◽  
Low Temperatures ◽  
Pressure Range ◽  
Low Pressure ◽  
Temperature Range ◽  
Inversion Temperature ◽  
Pressure Steam ◽  
Correlating Equations ◽  
First Time

Equations specifying the dynamic viscosity of compressed water and steam are presented. In the temperature range 0-100cC the location of the inversion locus (mu) is defined for the first time with some precision. The low pressure steam results are re-correlated and a higher inversion temperature is indicated than that previously accepted. From 100 to 600°C values of viscosity are derived up to 3·5 kilobar and between 600 and 1500°C up to 1 kilobar. All the original observations in the gaseous phase have been corrected to a consistent set of densities and deviation plots for all the new correlations are given. Although the equations give values within the tolerances of the International Skeleton Table it is clear that the range and tolerances of the latter could with some advantage be revised to give twice the existing temperature range and over 10 times the existing pressure range at low temperatures. A list of the observations used and their deviations from the correlating equations is available as a separate publication.

Unprecedented CO2 adsorption behaviour by 5A-type zeolite discovered in lower pressure region and at 300 K

2021 ◽  
Author(s):  
Akira Oda ◽  
Suguru Hiraki ◽  
Eiji Harada ◽  
Ikuka Kobayashi ◽  
Takahiro Ohkubo ◽  
...  
Keyword(s):  
Pressure Range ◽  
Co2 Adsorption ◽  
Low Pressure ◽  
Lower Pressure ◽  
Dft Studies ◽  
Adsorption Behaviour ◽  
Zeolite Sample ◽  
Pressure Region ◽  
Type Zeolite ◽  
Efficient Adsorbent

The NaCaA-85 zeolite sample which works as an efficient adsorbent for CO2 at RT and in low pressure range was found and its specificity is nicely explained by the model composed of CO2 pinned by two types of Ca2+ ions through far-IR and DFT studies.

scholarly journals An ICR Study of an Association Reaction at Low Pressure

1980 ◽  
Vol 87 ◽  
pp. 305-306
Author(s):  
M.J. Mcewan ◽  
V. G. Anicich ◽  
W.T. Huntress ◽  
P. R. Kemperer ◽  
M. T. Bowers
Keyword(s):  
Pressure Range ◽  
Rate Coefficient ◽  
Low Pressure ◽  
Second Order ◽  
Radiative Association ◽  
Association Reaction

An ICR investigation of the association reactionCH3+ + HCN CH3.HCN+has shown the reaction follows second order kinetics over the pressure range 1 × 10-6 to 3 × 10-4 Torr with a rate coefficient of 2 × 10-10 cm3s-1. These results can be interpreted in terms of a saturated 3-body or radiative association mechanism.

Application of Pressure-Sensitive Paints to Low-Pressure Range

2005 ◽  
Vol 19 (1) ◽  
pp. 9-16 ◽  
Author(s):  
Tomohide Niimi ◽  
Masaki Yoshida ◽  
Makoto Kondo ◽  
Yusuke Oshima ◽  
Hideo Mori ◽  
...  
Keyword(s):  
Pressure Range ◽  
Low Pressure ◽  
Pressure Sensitive

Calculation studies of coated particles performance in sodium-cooled fast reactor

Kerntechnik ◽  
2021 ◽  
Vol 86 (1) ◽  
pp. 45-49
Author(s):  
N. V. Maslov ◽  
E. I. Grishanin ◽  
P. N. Alekseev
Keyword(s):  
Fast Reactor ◽  
Reactor Core ◽  
Heat Removal ◽  
Design Solution ◽  
Fast Neutrons ◽  
Steel Shell ◽  
Primary Circuit ◽  
Coated Particles ◽  
The Core ◽  
Fuel Assemblies

Abstract This paper presents results of calculation studies of the viability of coated particles in the conditions of the reactor core on fast neutrons with sodium cooling, justifying the development of the concept of the reactor BN with microspherical fuel. Traditional rod fuel assemblies with pellet MOX fuel in the core of a fast sodium reactor are directly replaced by fuel assemblies with micro-spherical mixed (U,Pu)C-fuel. Due to the fact that the micro-spherical (U, Pu)C fuel has a developed heat removal surface and that the design solution for the fuel assembly with coated particles is horizontal cooling of the microspherical fuel, the core has additional possibilities of increasing inherent (passive) safety and improve the competitiveness of BN type of reactors. It is obvious from obtained results that the microspherical (U, Pu)C fuel is limited with the maximal burn-up depth of ∼11% of heavy atoms in conditions of the sodium-cooled fast reactor core at the conservative approach; it gives the possibility of reaching stated thermal-hydraulic and neutron-physical characteristics. Such a tolerant fuel makes it less likely that fission products will enter the primary circuit in case of accidents with loss of coolant and the introduction of positive reactivity, since the coating of microspherical fuel withstands higher temperatures than the steel shell of traditional rod-type fuel elements.

Buoyancy Effect on Heat Transfer in 5×5 Rod Bundles

2017 ◽  
Author(s):  
Da Liu ◽  
Fujun Gan ◽  
Chaozhu Zhang ◽  
Hanyang Gu
Keyword(s):  
Heat Transfer ◽  
Experimental Data ◽  
Mixed Convection ◽  
Buoyancy Force ◽  
Reynold Number ◽  
Low Flow ◽  
Buoyancy Effect ◽  
Rod Bundles ◽  
Convection Regime ◽  
Dc Power

Experiments of heat transfer at low flow rate are performed in a 5×5 square arrayed rod bundles. The diameter of the rod is 10mm with a pitch of 13.3mm, length of the test section is about 3 meters. Inlet Reynold number ranges from 2000 to 30000, Bo * ranges from 4×10−6 to 5×10−3. The rods are heated using a DC power, the heat flux ranges from 30 to 300 kW/m2. The experiment is aimed to investigate the buoyancy effect of mixed convection in rod bundles. The experimental data shows that similar with mixed convection in circular channels, buoyancy force has great effect on heat transfer at mixed convection regime in rod bundles. But the buoyancy effect appears at higher Bo* conditions. The spacer effect have also been investigated at both turbulent forced convection regime and mixed convection regime. The reconstruction of heat transfer downstream of spacers is different at different flow regimes, a reasonable explanation was provided.

scholarly journals Research on Influence of Different Simulation Methods of Bypass Flow in Thermal Hydraulic Analysis on Temperature Distribution in HTR-10

2020 ◽  
Vol 2020 ◽  
pp. 1-8
Author(s):  
Shiyan Sun ◽  
Youjie Zhang ◽  
Yanhua Zheng
Keyword(s):  
Temperature Distribution ◽  
Reactor Core ◽  
Flow Distribution ◽  
Maximum Temperature ◽  
Flow Ratio ◽  
Hydraulic Analysis ◽  
Bypass Flow ◽  
Pebble Bed ◽  
The Core ◽  
Thermal Hydraulic Analysis

In pebble-bed high temperature gas-cooled reactor, gaps widely exist between graphite blocks and carbon bricks in the reactor core vessel. The bypass helium flowing through the gaps affects the flow distribution of the core and weakens the effective cooling of the core by helium, which in turn affects the temperature distribution and the safety features of the reactor. In this paper, the thermal hydraulic analysis models of HTR-10 with bypass flow channels simulated at different positions are designed based on the flow distribution scheme of the original core models and combined with the actual position of the core bypass flow. The results show that the bypass coolant flowing through the reflectors enhances the heat transfer of the nearby components efficiently. The temperature of the side reflectors and the carbon bricks is much lower with more side bypass coolant. The temperature distribution of the central region in the pebble bed is affected by the bypass flow positions slightly, while that of the peripheral area is affected significantly. The maximum temperature of the helium, the surface, and center of the fuel elements rises as the bypass flow ratio becomes larger, while the temperature difference between them almost keeps constant. When the flow ratio of each part keeps constant, the maximum temperature almost does not change with different bypass flow positions.

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