Design of a Small Modular Nuclear Reactor with dual cooled annular fuel and investigation of the fuel inner radius effect on the power peaking factor and natural circulation parameters

2020 ◽  
Vol 138 ◽  
pp. 107185
Author(s):  
M. Zaidabadi Nejad ◽  
G.R. Ansarifar
2014 ◽  
Vol 986-987 ◽  
pp. 231-234
Author(s):  
Jun Teng Liu ◽  
Qi Cai ◽  
Xia Xin Cao

This paper regarded CNP1000 power plant system as the research object, which is the second-generation half Nuclear Reactor System in our country, and tried to set Westinghouse AP1000 passive residual heat removal system to the primary circuit of CNP1000. Then set up a simulation model based on RELAP5/MOD3.2 program to calculate and analyze the response and operating characteristic of passive residual heat removal system on assumption that Station Blackout occurs. The calculation has the following conclusions: natural circulation was quickly established after accident, which removes core residual heat effectively and keep the core safe. The residual heat can be quickly removed, and during this process the actual temperature was lower than saturation temperature in reactor core.


2020 ◽  
Vol 328 ◽  
pp. 01009
Author(s):  
František Világi ◽  
Branislav Knížat ◽  
Róbert Olšiak ◽  
Marek Mlkvik ◽  
Peter Mlynár ◽  
...  

The natural circulation helium loop is an experimental facility designed for the research of the possibility of utilizing natural convection cooling for the case of decay heat removal from a fast nuclear reactor. This concept would bring an improved automated safety system for future nuclear power plants operating a gas-cooled reactor. The article presents a new possibility of direct use of energy conservation laws in a 1D simulation of natural circulation loops. The calculation is performed by a triple iteration process, nested into each other. The results of the calculations showed good agreement with the measurements at steady state. A calculation with the proposed model at unsteady state is not yet possible, especially because of the exclusion of heat accumulation into the material.


Author(s):  
A. L. Laursen ◽  
F. J. Moody ◽  
J. C. Law

Spent nuclear fuel is currently being stored at nuclear reactor sites. The spent fuel removed from the reactor is first placed in a large water pool to remove the initial decay heat. After several years, when the decay heat has dropped below a set level, the fuel is moved into concrete storage casks where natural circulation continues the cooling process. The purpose of this report is to predict, using a simplified analysis, how hot the fuel rods get when cooled by air in the cask. The increase in temperature and the decrease in density cause a chimney effect in the cask. This paper presents an analytical method of obtaining maximum fuel clad temperature in the cask. A non-dimensional model is derived, which is used to calculate the entrance and exit air velocities of the cask. The relationship between these velocities and the temperature used to obtain the maximum fuel clad temperature. A numerical scheme used to predict the maximum temperature is presented here and the results are compared to the analytical model. Both methods yielded corroborating results for fuel placed in the casks after spending similar amounts of time in a spent fuel pool.


Author(s):  
Tatiana Farkas ◽  
Iva´n To´th

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.


Author(s):  
Jeffrey Samuel ◽  
Glenn Harvel ◽  
Igor Pioro

The feasibility of operating with natural circulation as the normal mode of core cooling has been successfully demonstrated for a few small sized nuclear reactors. Natural circulation is being considered for cooling the core of a nuclear reactor under normal operating conditions in several advanced reactor concepts being developed today. Although studies have been conducted in natural circulation for many decades, using natural circulation as the primary cooling mechanism for nuclear reactors or as a passive safety system requires a comprehensive understanding of local and integral system phenomena, validated benchmark data, accurate predictive tools, and reliability analysis methods. As full-scale experiments of supercritical water are expensive, scaling laws can be applied to develop test matrices using modelling fluids to reproduce similar conditions in a scaled-down experimental thermalhydraulic loop. The main aim of this work is to understand the natural circulation phenomena by analyzing water and modelling fluids such as Carbon dioxide (CO2) and Freon 134a (R-134a). The use of the modelling fluids at subcritical, pseudocritical and supercritical pressures is discussed along with fluid-to-fluid scaling techniques. The results from a one-dimensional numerical model developed using MATLAB to calculate the steady-state mass flow rate and heat transport characteristics of an experimental natural circulation test loop are presented and analyzed.


1982 ◽  
Vol 53 (4) ◽  
pp. 720-724
Author(s):  
V. A. Blinkin ◽  
E. I. Emel'yanov

2008 ◽  
Vol 2008 ◽  
pp. 1-1
Author(s):  
Dilip Saha ◽  
John Cleveland

2021 ◽  
Vol 2072 (1) ◽  
pp. 012012
Author(s):  
R Wulandari ◽  
S Permana ◽  
Suprijadi

Abstract Natural convention, the heat transfer on fluid due to density differences that can be caused by differences in fluid temperature. One example application of natural convection is cooling system, such as nuclear reactor cooling system. The purpose of this study is to analysis the basic characteristic heat transfer of sodium liquid in the natural circulation system for steady state analysis and transient characteristic with Finite Element Method. The selected module is the Non-Isothermal FLow (NITF) module. This module is a combination of three basic equations, namely the continuity equation, the Navier-Stokes equation, and the dynamic equation of heat transfer in fluid. The simulation model measures 1.5 x 2 (m) with sodium liquid (Na) as a fluid.


Author(s):  
Rong Cai ◽  
Nina Yue ◽  
Hongyu Fang ◽  
Baowen Chen ◽  
Lili Liu ◽  
...  

Abstract The marine nuclear power plant operating in the marine environment has complicated motion under the influence of wind and waves. The movement of marine nuclear power plant will affect the thermal-hydraulic characteristics of its nuclear reactor system. Compared with other typical motion conditions, the effects of rolling conditions on the thermal-hydraulic characteristics of the nuclear reactor system are the most complex. In order to study the effects of rolling conditions on the thermal-hydraulic characteristics of the nuclear reactor system, a thermal-hydraulic system code for motion conditions named STAC was developed. The STAC code was verified by the experiments conducted in Japan. The effects of rolling conditions on the thermal-hydraulic characteristics of the nuclear reactor system under forced circulation and natural circulation are studied with the STAC code. The simulation results show that the thermal parameters of the reactor system under rolling condition fluctuate periodically. The fluctuation period of the thermal parameters of the core is half of the rolling period, and the fluctuation periods of other thermal parameters are the same as the rolling period. The effect of rolling condition on thermal-hydraulic parameters under forced circulation is smaller than that under natural circulation. The fluctuation amplitudes of the thermal parameters increase with the angle amplitude of the rolling condition. There is a rolling period with the smallest fluctuation amplitude. Under the rolling condition with short period, the fluctuation amplitudes of the thermal parameters increase and their average values change rapidly as the rolling period decreases. Under the rolling condition with large period, the fluctuation amplitudes of the thermal parameters increase with the rolling period, and they tend to fixed values.


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