Thermal-hydraulic modeling and acoustic correlation (Conference Presentation)

2020 ◽  
Author(s):  
Vincent Bruyere ◽  
Nicolas Authier ◽  
Patrick Namy
Keyword(s):  
Hydraulic Modeling ◽  
Thermal Hydraulic Modeling

Thermal–hydraulic modeling and transient analysis of helium-cooled tungsten target coupled with ADS core

2016 ◽  
Vol 87 ◽  
pp. 612-620
Author(s):  
Tianji Peng ◽  
Zhiwei Zhou ◽  
Sicong Xiao ◽  
Xuanyu Sheng ◽  
Long Gu
Keyword(s):  
Transient Analysis ◽  
Hydraulic Modeling ◽  
Tungsten Target ◽  
Thermal Hydraulic Modeling

Thermal Hydraulic Analysis of Severe Accident in PFBR

2010 ◽  
Author(s):  
K. Velusamy ◽  
P. Chellapandi ◽  
G. R. Raviprasan ◽  
P. Selvaraj ◽  
S. C. Chetal
Keyword(s):  
Atmospheric Pressure ◽  
Fluid Dynamic ◽  
Hydraulic Modeling ◽  
Severe Accident ◽  
Maximum Temperature ◽  
Maximum Pressure ◽  
Dynamic Calculation ◽  
Thermal Hydraulic Modeling ◽  
Thermal Hydraulic Analysis ◽  
Design Pressure

During a core disruptive accident (CDA), the amount of primary sodium that can be released to Reactor Containment Building (RCB) in Prototype Fast Breeder Reactor (PFBR) is estimated to be 350 kg/s, by a transient fluid dynamic calculation. The pressure and temperature evolutions inside RCB, due to consequent sodium fire have been estimated by a constant burning rate model, accounting for heat absorption by RCB wall, assuming RCB isolation based on area gamma monitors. The maximum pressure developed is 7000 Pa. In case RCB isolation is delayed, then the final pressure inside RCB reduces below atmospheric pressure due to cooling of RCB air. The negative pressure that can be developed is estimated by dynamic thermal hydraulic modeling of RCB air / wall to be −3500 Pa. These investigations were useful to arrive at the RCB design pressure. Following CDA, RCB is isolated for 40 days. During this period, the heat added to RCB is dissipated to atmosphere only by natural convection. Considering all the possible routes of heat addition to RCB, evolution of RCB wall temperature has been predicted using HEATING5 code. It is established that the maximum temperature in RCB wall is less than the permissible value.

Thermal-hydraulic modeling and analysis of spool valve with sloping U-shape notch by bond graph

2015 ◽  
Vol 22 (11) ◽  
pp. 4205-4212
Author(s):  
Lei Lou ◽  
Wan-rong Wu ◽  
Zhao-Qiang Wang ◽  
Xiang-jing Liang
Keyword(s):  
Hydraulic Modeling ◽  
Bond Graph ◽  
Modeling And Analysis ◽  
Thermal Hydraulic Modeling ◽  
Spool Valve

Simulation of Nuclear Fuel Behavior in Accident Conditions With the DIONISIO Code

2019 ◽  
Vol 5 (2) ◽  
Author(s):  
Martín Lemes ◽  
Alicia Denis ◽  
Alejandro Soba
Keyword(s):  
Nuclear Fuel ◽  
Computer Code ◽  
Hydraulic Modeling ◽  
Application Range ◽  
Behavior Simulation ◽  
External Conditions ◽  
Capture And Release ◽  
Fuel Behavior ◽  
Accident Conditions ◽  
Thermal Hydraulic Modeling

DIONISIO is a computer code designed to simulate the behavior of one nuclear fuel rod during its permanence within the reactor. Starting from the power history and the external conditions to which the rod is subjected, the code predicts all the meaningful variables of the system. Its application range has been recently extended to include accidental conditions, in particular the so-called loss of coolant accidents (LOCA). In order to make realistic predictions, the conditions in the rod environment have been taken into account since they represent the boundary conditions with which the differential equations describing the fuel phenomena are solved. Without going into the details of the thermal-hydraulic modeling, which is the task of the specific codes, a simplified description of the conditions in the cooling channel during a LOCA event has been developed and incorporated as a subroutine of DIONISIO. This has led to an improvement of the fuel behavior simulation, which is evidenced by the considerable number of comparisons with experiments carried out, many of them reported in this paper. Moreover, this work describes a model of high temperature capture and release of hydrogen in the nuclear fuel cladding, in scenarios typical of LOCA events. The corresponding computational model is being separately tested and will be next included in the DIONISIO thermal-hydraulic module.

Thermal-hydraulic modeling of reactivity accident in MTR reactors

2007 ◽  
Vol 34 (9) ◽  
pp. 732-742 ◽  
Author(s):  
Hany Khater ◽  
Talal Abu-EL-Maty ◽  
Salah El-Din EL-Morshdy
Keyword(s):  
Hydraulic Modeling ◽  
Thermal Hydraulic Modeling ◽  
Reactivity Accident

Thermal-Hydraulic Modeling of Supercritical Water Oxidation of Ethanol

1996 ◽  
Vol 10 (2) ◽  
pp. 326-332 ◽  
Author(s):  
Chang H. Oh ◽  
Robert J. Kochan ◽  
Thomas R. Charlton ◽  
Alain L. Bourhis
Keyword(s):  
Supercritical Water ◽  
Water Oxidation ◽  
Hydraulic Modeling ◽  
Supercritical Water Oxidation ◽  
Thermal Hydraulic Modeling ◽  
Oxidation Of Ethanol
Sign in / Sign up

Export Citation Format

Share Document