Experimental Studies on the Use of Liquid Lead in a Molten Salt Nuclear Reactor

1983 ◽  
Vol 63 (2) ◽  
pp. 197-208 ◽  
Author(s):  
M. Broc ◽  
J. Sannier ◽  
G. Santarini
Keyword(s):  
Molten Salt ◽  
Nuclear Reactor ◽  
Experimental Studies ◽  
Liquid Lead

Coupled Analysis of Thermal Hydraulics and Neutronics for a Molten Salt Reactor

2017 ◽  
Author(s):  
Jian Ge ◽  
Dalin Zhang ◽  
Wenxi Tian ◽  
Suizheng Qiu ◽  
G. H. Su
Keyword(s):  
Molten Salt ◽  
Nuclear Reactor ◽  
Coupled Analysis ◽  
Molten Salt Reactor ◽  
Thermal Hydraulics ◽  
Delayed Neutron ◽  
Analysis Model ◽  
Liquid Salt ◽  
Energy Equations ◽  
Volume Method

As one of the six selected optional innovative nuclear reactor in the generation IV International Forum (GIF), the Molten Salt Reactor (MSR) adopts liquid salt as nuclear fuel and coolant, which makes the characteristics of thermal hydraulics and neutronics strongly intertwined. Coupling analysis of neutronics and thermal hydraulics has received considerable attention in recent years. In this paper, a new coupling method is introduced based on the Finite Volume Method (FVM), which is widely used in the Computational Fluid Dynamics (CFD) methodology. Neutron diffusion equations and delayed neutron precursors balance equations are discretized and solved by the commercial CFD package FLUENT, along with continuity, momentum and energy equations simultaneously. A Temporal And Spatial Neutronics Analysis Model (TASNAM) is developed using the User Defined Functions (UDF) and User Defined Scalar (UDS) in FLUENT. A neutronics benchmark is adopted to demonstrate the solution capability for neutronics problems using the method above. Furthermore, a steady state coupled analysis of neutronics and thermal hydraulics for the Molten Salt Advanced Reactor Transmuter (MOSART) is performed. Two groups of neutrons and six groups of delayed neutron precursors are adopted. Distributions of the liquid salt velocity, temperature, neutron flux and delayed neutron precursors in the core are obtained and analyzed. This work can provide some valuable information for the design and research of MSRs.

scholarly journals Thermal Hydraulics Analysis of the Distribution Zone in Small Modular Dual Fluid Reactor

Metals ◽  
2020 ◽  
Vol 10 (8) ◽  
pp. 1065
Author(s):  
Chunyu Liu ◽  
Xiaodong Li ◽  
Run Luo ◽  
Rafael Macian-Juan
Keyword(s):  
Heat Transfer ◽  
Molten Salt ◽  
Heat Removal ◽  
Core Region ◽  
Major Influence ◽  
Liquid Lead ◽  
Thermal Hydraulics ◽  
The Core ◽  
Removal Capacity ◽  
Power Peak

The Small Modular Dual Fluid Reactor (SMDFR) is a novel molten salt reactor based on the dual fluid reactor concept, which employs molten salt as fuel and liquid lead/lead-bismuth eutectic (LBE) as coolant. A unique design of this reactor is the distribution zone, which locates under the core and joins the core region with the inlet pipes of molten salt and coolant. Since the distribution zone has a major influence on the heat removal capacity in the core region, the thermal hydraulics characteristics of the distribution zone have to be investigated. This paper focuses on the thermal hydraulics analysis of the distribution zone, which is conducted by the numerical simulation using COMSOL Multiphysics with the CFD (Computational Fluid Dynamics) module and the Heat Transfer module. The energy loss and heat exchange in the distribution zone are also quantitatively analyzed. The velocity and temperature distributions of both molten salt and coolant at the outlet of the distribution zone, as inlet of the core region, are produced. It can be observed that the outlet velocity profiles are proportional in magnitude to the inlet velocity ones with a similar shape. In addition, the results show that the heat transfer in the center region is enhanced due to the velocity distribution, which could compensate the power peak and flatten the temperature distribution for a higher power density.

A Numerical Tool for Space Nuclear Reactor Design based on Molten Salt Reactors (MSRs)

2020 ◽  
Author(s):  
J. Blanco ◽  
V. Ghetta ◽  
J. Giraud ◽  
V. Richard ◽  
P. Rubiolo ◽  
...  
Keyword(s):  
Molten Salt ◽  
Nuclear Reactor ◽  
Reactor Design ◽  
Numerical Tool

Preliminary Study of the Use of Freeze-Valves for a Passive Shutdown System in Molten Salt Reactors

2014 ◽  
Author(s):  
Qiming Li ◽  
Zhongfeng Tang ◽  
Yuan Fu ◽  
Zhong Li ◽  
Naxiu Wang
Keyword(s):  
Molten Salt ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Mechanical Valve ◽  
Thermal Design ◽  
Normal Operation ◽  
Passive System ◽  
Fuel Salt ◽  
Preliminary Study ◽  
Accident Conditions

The use of passive shutdown systems to enhance safety is one element of next-generation reactor design. The Freeze-Valve has been proposed as a key device in the passive system to stop the chain reaction of the Molten Salt Reactor (MSR), which has been chosen by Generation IV International Forum (GIF) as one of the six Generation IV reactor concepts. During reactor normal operation, the molten salt in the valve is cooled to a solid plug. In the event that the reactor overheats under accident conditions when all other active control systems fail, the plug will melt. The liquid fuel salt will be pulled out from the reactor core by gravity into dump tanks, and criticality will cease because the reaction is no longer moderated by the graphite in the reactor core. The more accurate the Freeze-Valve’s thermal design is, the more efficient the passive shutdown system becomes. In this study, an investigation of the thermal performance of the Freeze-Valve is conducted based on finite element methods verified by experimental data, and some modified designs are presented with recommendations. For further consideration, some innovative governing techniques used to control the Freeze-Valve are discussed in detail. Here, a more critical thermal design is focused on that can make the passive system shut down the nuclear reactor quickly and reliably. The Freeze-Valve can be used in the molten salt loop rather than a mechanical valve, which may become jammed by frozen salt. Paper published with permission.

scholarly journals Development of a cold plug valve with fluoride salt

2019 ◽  
Vol 5 ◽  
pp. 9 ◽  
Author(s):  
Julien Giraud ◽  
Veronique Ghetta ◽  
Pablo Rubiolo ◽  
Mauricio Tano Retamales
Keyword(s):  
Heat Transfer ◽  
Molten Salt ◽  
Electric Power ◽  
Heat Storage ◽  
Experimental Studies ◽  
Molten Salt Reactor ◽  
Fluoride Salt ◽  
Working Principle ◽  
Thermal Simulations ◽  
Plug Valve

Experimental studies have been developed on a new freeze plug concept for safety valves in facilities using molten salt. They are designed to allow the closure of an upstream circuit by solidifying the molten salt in a section of the device and to passively melt in case of a loss of electric power, thus releasing the upper fluid. The working principle of these cold plug designs relies on the control of the heat transfer balance inside the device, which determines whether the salt inside the cold plug solidifies or melts. The device is mainly composed of steel masses that are dimensioned to provide sufficient thermal heat storage to melt the salt and thus open the cold plug after the electric power is stopped. The final goal of the work is to provide useful recommendations and guidelines for the design of a cold plug for the emergency draining system of a molten salt reactor. Some numerical thermal simulations were performed with ANSYS mechanical (Finite Element Method) to be compared with results of the experiments and to make extrapolations for a new component to be used in a reactor.

A Numerical Tool for Space Nuclear Reactor Design based on Molten Salt Reactors (MSRs)

2020 ◽  
Author(s):  
J. Blanco ◽  
V. Ghetta ◽  
J. Giraud ◽  
V. Richard ◽  
P. Rubiolo ◽  
...  
Keyword(s):  
Molten Salt ◽  
Nuclear Reactor ◽  
Reactor Design ◽  
Numerical Tool

Development of Supercritical Water Heat-Transfer Correlation for Vertically-Upward Internally-Ribbed Tubes

2014 ◽  
Author(s):  
Weiqiang Zhang ◽  
Huixiong Li ◽  
Qing Zhang ◽  
Yifang Zhang ◽  
Tai Wang
Keyword(s):  
Heat Transfer ◽  
Experimental Data ◽  
Supercritical Water ◽  
Nuclear Reactor ◽  
Large Deviation ◽  
Experimental Studies ◽  
Great Influence ◽  
Dimensionless Numbers ◽  
Experimental Database ◽  
Heat Transfer Correlations

The investigation on the heat transfer characteristics for supercritical pressure water (SCW) is of value for the development of the supercritical water-cooled nuclear reactor (SCWR). As an important heat transfer enhancement element, heat transfer for SCW in internally-ribbed tubes was still not solved, though lots of experimental studies have been published and a great many heat transfer correlations were proposed. This paper presented an analysis of heat transfer in the internally-ribbed tubes, through comparing heat transfer correlations for SCW gained from different internally-ribbed tubes under the same operating condition. It was found that all existing heat transfer correlations reported could not been well applied for various internally-ribbed tubes with large deviation between prediction results and experimental values, because rib geometry had a great influence on heat transfer of internally-ribbed tubes. On the basis of experimental data collected from open literature for internally-ribbed tubes, a new general calculation correlation of heat transfer coefficient for SCW was developed for various internally-ribbed tubes by combining an optimized empirical correlation for vertically-upward smooth tubes and four dimensionless numbers of rib geometry. The results show that the calculated values of the new present correlation is in reasonable agreement with available experimental data collected. Moreover, the new correlation was verified well by experiment data of two new-type internally-ribbed tubes performed beyond the above experimental database.

Sign in / Sign up

Export Citation Format

Share Document