Study on the Coupled Neutronic and Thermal-Hydraulic Characteristics of the New Concept Molten Salt Reactor

2009 ◽  
Author(s):  
Peng Wang ◽  
Suizheng Qiu
Keyword(s):  
Molten Salt ◽  
Flow Velocity ◽  
Design Theory ◽  
Primary Circuit ◽  
Molten Salt Reactor ◽  
Hydraulic Characteristics ◽  
Delayed Neutron ◽  
Delayed Neutrons ◽  
Flow Effect ◽  
Fuel Salt

The new concept Molten Salt Reactor is the only liquid-fuel reactor of the six Generation IV advanced nuclear energy systems. The liquid molten salt serves as the fuel and coolant simultaneously and causes one important feature: the delayed neutrons precursors are drifted by the fuel flow, which leads the spread of delayed neutrons distribution to non-core parts of the primary circuit, and it also can result in a reactivity variation depending on the flow condition of the fuel salt. Therefore, the neutronic and thermal-hydraulic characteristics of the Molten Salt Reactor is quite different from the conventional nuclear reactors using solid fissile materials, and no other reactor design theory and safety analysis methodologies can be used for reference. The neutronic model is derived based on the conservation of particle considering the flow effect of the fuel salt in the Molten Salt Reactor, while the thermal-hydraulic model uses the fundamental conservation laws: the mass, momentum and energy conservation equations. Then the neutronic and thermal-hydraulic calculations were coupled and the influences of inflow temperature and flow velocity on the reactor physical properties were obtained. The calculated results show that the flow effect on the distributions of thermal and fast neutron fluxes is very weak, as well as on the effective multiplication factor keff. While the flow effect on the distribution of delayed neutron precursors is much stronger. The inflow temperature influences the distribution of neutron flux and delayed neutron precursors slightly, and makes significant negative reactivity. Coupled calculation also reveals that the flow velocity of molten salt has little effect on the distribution of neutron fluxes in the steady state, but affects the delayed neutron precursors’ distribution significantly.

Study on the Coupled Neutronic and Thermal-Hydraulic Characteristics of the New Concept Molten Salt Reactor

2010 ◽  
Vol 132 (10) ◽  
Author(s):  
Peng Wang ◽  
Libo Qian ◽  
Dalin Zhang ◽  
Wenxi Tian ◽  
Guanghui Su ◽  
...  
Keyword(s):  
Molten Salt ◽  
Flow Velocity ◽  
Design Theory ◽  
Primary Circuit ◽  
Molten Salt Reactor ◽  
Hydraulic Characteristics ◽  
Delayed Neutron ◽  
Flow Effect ◽  
Fuel Reactor ◽  
Fuel Salt

The new concept molten salt reactor is the only liquid-fuel reactor of the six Generation IV advanced nuclear energy systems. The liquid molten salt serves as the fuel and coolant simultaneously and causes one important feature: the delayed neutron precursors are drifted by the fuel flow, which leads the spread of delayed neutrons’ distribution to noncore parts of the primary circuit, and it also results in reactivity variation depending on the flow condition of the fuel salt. Therefore, the neutronic and thermal-hydraulic characteristics of the molten salt reactor are quite different from the conventional nuclear reactors using solid fissile materials. Besides, there is no other reactor design theory and safety analysis methodologies can be used for reference. The neutronic model is derived based on the conservation of particles considering the flow effect of the fuel salt in the molten salt reactor, while the thermal-hydraulic model applies the fundamental conservation laws: the mass, momentum, and energy conservation equations. Then, the neutronic and thermal-hydraulic calculations are coupled and the influences of inflow temperature and flow velocity on the reactor physical properties are obtained. The calculated results show that the flow effect on the distributions of thermal and fast neutron fluxes is very weak, as well as on the effective multiplication factor keff, while the flow effect on the distribution of delayed neutron precursors is much stronger. The inflow temperature influences the distribution of neutron fluxes and delayed neutron precursors slightly, and makes a significant negative reactivity. Coupled calculation also reveals that the flow velocity of molten salt has little effect on the distribution of neutron fluxes in the steady-state, but affects the delayed neutron precursors’ distribution significantly.

Reactor Controllability of 3-Region-Core Molten Salt Reactor System: A Study on Load Following Capability

2006 ◽  
Author(s):  
Takahisa Yamamoto ◽  
Koshi Mitachi ◽  
Masatoshi Nishio
Keyword(s):  
Molten Salt ◽  
Solid Fuel ◽  
Molten Salt Reactor ◽  
Delayed Neutron ◽  
Reactor Kinetics ◽  
External Loop ◽  
Fuel Salt ◽  
Salt Flow ◽  
Point Reactor Kinetics ◽  
Graphite Moderator

The Molten Salt Reactor (MSR) systems are liquid-fueled reactors that can be used for actinide burning, production of electricity, production of hydrogen, and production of ssile fuels (breeding). Thorium (Th) and uranium-233 (233U) are fertile and ssile of the MSR systems, and dissolved in a high-temperature molten fluoride salt (fuel salt) with a very high boiling temperature (up to 1650K), that is both the reactor nuclear fuel and the coolant. The MSR system is one of the six advanced reactor concepts identified by the Generation IV International Forum (GIF) as a candidate for cooperative development [1]. In the MSR system, fuel salt flows through a fuel duct constructed around a reactor core and fuel channel of a graphite moderator accompanied by fission reaction and heat generation, and flows out to an external-loop system consisted of a heat exchanger and a circulation pump. Due to the motion of fuel salt, delayed neutron precursors that are one of the source of neutron production make to change their position between the ssion reaction and neutron emission events and decay even occur in the external loop system. Hence the reactivity and effective delayed neutron precursor fraction of the MSR system are lower than those of solid fuel reactor systems such as Boiling Water Reactors (BWRs) and Pressurised Water Reactor (PWRs). Since all of the presently operating nuclear power reactors utilize solid fuel, little attention had been paid to the MSR analysis of the reactivity loss and reactor characteristics change caused by the fuel salt circulation. Sides et al. [2] and Shimazu et al. [3] developed MSR analytical models based on the point reactor kinetics model to consider the effect of fuel salt flow. Their models represented a reactor as having six zones for fuel salt and three zones for the graphite moderator. Since their models employed the point reactor kinetics model and the rough temperature approximation, their results were not sufficiently accurate to consider the effect of fuel salt flow.

Numerical Research on Steady Coupling of Neutronics and Thermal-Hydraulics for a Molten Salt Reactor

2008 ◽  
Author(s):  
Dalin Zhang ◽  
Changliang Liu ◽  
Libo Qian ◽  
Guanghui Su ◽  
Suizheng Qiu
Keyword(s):  
Thermal Neutron ◽  
Molten Salt ◽  
Molten Salt Reactor ◽  
Thermal Hydraulics ◽  
Steady Condition ◽  
Delayed Neutron ◽  
The Core ◽  
Production Of Hydrogen ◽  
Fuel Salt ◽  
Salt Flow

The Molten Salt Reactor (MSR), which is one of the ‘Generation IV’ concepts, can be used for production of electricity, actinide burning, production of hydrogen, and production of fissile fuels. In this paper, a single-liquid-fueled MSR was selected for conceptual research. For this MSR, a ternary system of 15%LiF-58%NaF-27%BeF2 was proposed as the reactor fuel solvent, coolant and also moderator with ca. 1 mol% UF4 dissolving in it, which circulates through the whole primary loop accompanying fission reaction only in the core. The fuel salt flow makes the MSR different from the conventional reactors using solid fissile materials, and makes the neutronics and thermal-hydraulic coupled strongly, which plays the important role in the research of reactor safety analysis. Therefore, it’s necessary to study the coupling of neutronics and thermal-hydraulic. The theoretical models of neutronics and thermal-hydraulics under steady condition were conducted and calculated by numerical method in this paper. The neutronics model consists of two group neutron diffusion equations for fast and thermal neutron fluxes, and balance equations for six-group delayed neutron precursors considering flow effect. The thermal-hydraulic model was founded on the base of the fundamental conservation laws: the mass, momentum and energy conservation equations. These two models were coupled through the temperature and heat source. The spatial discretization of the above models is based on the finite volume method (FVM), and the thermal-hydraulic equations are computed by SIMPLER algorithm with domain extension method on the staggered grid system. The distribution of neutron fluxes, the distribution of the temperature and velocity and the distribution of the delayed neutron precursors in the core were obtained. The numerical calculated results show that, the fuel salt flow has little effect to the distribution of fast and thermal neutron fluxes and effective multiplication factor; however, it affects the distribution of the delayed neutron precursors significantly, especially long-lived one. In addition, it could be found that the delayed neutron precursors influence the neutronics slightly under the steady condition, and the flow could remove the heat generated by the neutron reactions easily to ensure the reactor safe. The obtained results serve some valuable information for the research and design of this new generation reactor.

Steady state investigation on neutronics of a molten salt reactor considering the flow effect of fuel salt

2008 ◽  
Vol 32 (8) ◽  
pp. 624-628 ◽  
Author(s):  
Zhang Da-Lin ◽  
Qiu Sui-Zheng ◽  
Liu Chang-Liang ◽  
Su Guang-Hui
Keyword(s):  
Steady State ◽  
Molten Salt ◽  
Molten Salt Reactor ◽  
Flow Effect ◽  
Fuel Salt

Steady State Analysis of Molten Salt Reactor in Consideration of the Effect of Fuel Salt Flow

2004 ◽  
Author(s):  
Takahisa Yamamoto ◽  
Koshi Mitachi ◽  
Takashi Suzuki
Keyword(s):  
Steady State ◽  
Thermal Neutron ◽  
Molten Salt ◽  
Fuel Cycle ◽  
Molten Salt Reactor ◽  
Delayed Neutron ◽  
Steady State Analysis ◽  
Fuel Salt ◽  
Salt Flow ◽  
State Analysis

The Molten Salt Reactor (MSR) is a thermal neutron reactor with graphite moderation and operates on the thorium-uranium fuel cycle. The feature of the MSR is that fuel salt flows the inside of the reactor accompanying nuclear fission reaction. In the previous study, the authors had developed numerical model to simulate the effects of the fuel salt flow on the reactor characteristics. This paper applies the model to the steady state analysis of the small MSR system and estimates the effects of the fuel flow. The model consists of two group diffusion equations for fast and thermal neutron fluxes, balance equations for six-group delayed neutron precursors and energy conservation equations for fuel salt and graphite moderator. The following results are obtained: (1) the fuel salt flow affects the distributions of the delayed neutron precursors, especially long-lived one, and (2) the extension of residence time in the external loop system and the rise of fuel inflow temperature slightly show negative reactivity effects, decreasing neutron multiplication factor of the small MSR system.

A new method for monitoring the redox potential of fuel salt based on the deposition of 95Nb on Hastelloy C276

2021 ◽  
Vol 109 (5) ◽  
pp. 357-365
Author(s):  
Zhiqiang Cheng ◽  
Zhongqi Zhao ◽  
Junxia Geng ◽  
Xiaohe Wang ◽  
Jifeng Hu ◽  
...  
Keyword(s):  
Redox Potential ◽  
Molten Salt ◽  
Specific Activity ◽  
Chemical Reduction ◽  
Quantitative Approach ◽  
Experimental Results ◽  
New Method ◽  
Molten Salt Reactor ◽  
Fuel Salt ◽  
Well Correlation

Abstract To develop the application of 95Nb as an indicator of redox potential for fuel salt in molten salt reactor (MSR), the specific activity of 95Nb in FLiBe salt and its deposition of 95Nb on Hastelloy C276 have been studied. Experimental results indicated that the amount of 95Nb deposited on Hastelloy C276 resulted from its chemical reduction exhibited a positive correlation with the decrease of 95Nb activity in FLiBe salt and the relative deposition coefficient of 95Nb to 103Ru appeared a well correlation with 95Nb activity in FLiBe salt. Both correlations implied that the measurement of 95Nb activity deposited on Hastelloy C276 specimen might provide a quantitative approach for monitoring the redox potential of fuel salt in MSR.

Flow effect on 135 I and 135 Xe evolution behavior in a molten salt reactor

2017 ◽  
Vol 314 ◽  
pp. 318-325 ◽  
Author(s):  
Jianhui Wu ◽  
Chen Guo ◽  
Xiangzhou Cai ◽  
Chenggang Yu ◽  
Chunyan Zou ◽  
...  
Keyword(s):  
Molten Salt ◽  
Molten Salt Reactor ◽  
Flow Effect ◽  
Evolution Behavior

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.

Numerical Analysis for a Molten Salt Reactor in the Presence of Fissile Lump

2012 ◽  
Author(s):  
Yao Xiao ◽  
Dalin Zhang ◽  
Zhangpeng Guo ◽  
Suizheng Qiu
Keyword(s):  
Molten Salt ◽  
Diffusion Theory ◽  
Neutron Diffusion ◽  
Delayed Neutron ◽  
Flow And Heat Transfer ◽  
Fuel Salt ◽  
Reactor Types ◽  
New Generation ◽  
Research And Design ◽  
Salt Flow

Molten salt reactors (MSRs) have seen a marked resurgence of interest over the past few decades, highlighted by their inclusion as one of the six Generation IV reactor types. The MSRs are characterized by using the fluid-fuel, so that their technologies are fundamentally different from those used in the conventional solid-fuel reactors. In this paper, the attention is focused on the behaviors of a MSR in the presence of localized perturbations caused by fissile precipitates. A neutron kinetic model considering the fuel salt flow is established based on the neutron diffusion theory, which consists of two-group neutron diffusion equations for the fast and thermal neutron fluxes and six-group balance equations for delayed neutron precursors, and the group constants dependent on the temperature are calculated by the code DRAGON. In addition, the k-epsilon turbulent model is adopted to establish the flow and heat transfer. The thermo-hydraulic and neutronic models which are coupled through the temperature, heat source and velocity are coded in a program. The effects of the localized perturbation on the distributions of power, temperature, neutron fluxes and delayed neutron precursors are obtained and discussed in detail. The results provide some valuable information for the research and design of this new generation reactor.

A Novel Molten Salt Reactor Concept to Implement the Multi-Step Time-Scheduled Transmutation Strategy

2002 ◽  
Author(s):  
Gyula Csom ◽  
Sandor Feher ◽  
Mate Szieberth
Keyword(s):  
Molten Salt ◽  
Liquid Fuel ◽  
Primary Circuit ◽  
Chemical Processing ◽  
Molten Salt Reactor ◽  
The Core ◽  
Continuous Feed ◽  
Fuel Mixtures ◽  
Continuous Chemical ◽  
Optimal Implementation

Nowadays the molten salt reactor (MSR) concept seems to revive as one of the most promising systems for the realization of transmutation. In the molten salt reactors and subcritical systems the fuel and material to be transmuted circulate dissolved in some molten salt. The main advantage of this reactor type is the possibility of the continuous feed and reprocessing of the fuel. In the present paper a novel molten salt reactor concept is introduced and its transmutational capabilities are studied. The goal is the development of a transmutational technique along with a device implementing it, which yield higher transmutational efficiencies than that of the known procedures and thus results in radioactive waste whose load on the environment is reduced both in magnitude and time length. The procedure is the multi-step time-scheduled transmutation, in which transformation is done in several consecutive steps of different neutron flux and spectrum. In the new MSR concept, named “multi-region” MSR (MRMSR), the primary circuit is made up of a few separate loops, in which salt-fuel mixtures of different compositions are circulated. The loop sections constituting the core region are only neutronically and thermally coupled. This new concept makes possible the utilization of the spatial dependence of spectrum as well as the advantageous features of liquid fuel such as the possibility of continuous chemical processing etc. In order to compare a “conventional” MSR and a proposed MRMSR in terms of efficiency, preliminary calculational results are shown. Further calculations in order to find the optimal implementation of this new concept and to emphasize its other advantageous features are going on.

Sign in / Sign up

Export Citation Format

Share Document