Calculation of delayed neutron precursors’ transit time in the external loop during a flow velocity transient in a Molten Salt Reactors

2022 ◽  
Vol 165 ◽  
pp. 108640
Author(s):  
Rodrigo Costa Diniz ◽  
Felipe Siqueira de Souza da Rosa ◽  
Alessandro da Cruz Gonçalves
Keyword(s):  
Molten Salt ◽  
Flow Velocity ◽  
Transit Time ◽  
Delayed Neutron ◽  
External Loop

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.

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.

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.

scholarly journals MULTIPHYSICS MODELING OF PRECURSORS IN MOLTEN SALT FAST REACTORS USING PROTEUS AND Nek5000

2021 ◽  
Vol 247 ◽  
pp. 06026
Author(s):  
B. Feng ◽  
T. Fei ◽  
D. Shaver ◽  
Y. Jung ◽  
J. Fang ◽  
...  
Keyword(s):  
Molten Salt ◽  
Discrete Ordinates ◽  
Delayed Neutron ◽  
Test Channel ◽  
Drift Model ◽  
Model Complex ◽  
Transient Simulations ◽  
Complex Core ◽  
And Performance ◽  
The Impact

The goal of this work was to calculate the impact of the delayed neutron precursor drift in fast spectrum Molten Salt Reactors (MSRs) using coupled solutions from the neutronics code PROTEUS and the computational fluid dynamics code Nek5000. Specifically, using a multiphysics approach to solve the effective delayed neutron fraction (βeff) or delayed neutron precursor distribution for reactors with flowing fuel salts would provide valuable information for transient simulations and safety assessments. Given the multiple options for the flux solution and geometric resolution/fidelity in PROTEUS, two approaches were developed and applied to various test cases: PROTEUS-NODAL/Nek5000 and PROTEUS-SN/Nek5000. For the former, the precursors are tracked in the built-in precursor drift model in PROTEUS-NODAL, whereas in the latter, Nek5000 directly tracks the precursors. Both approaches were used to solve a single test channel problem and showed excellent agreement in the calculated βeff. Separately, a 3D hourglass-shaped core was modeled using the PROTEUS-SN/Nek5000 approach. This problem was designed to demonstrate the capability of the discrete ordinates (SN) solver and Nek5000 to model complex core designs with axially varying geometries and the ability for Nek5000 to track the precursors and calculate the resulting βeff. In addition, the Nek5000 calculations revealed the presence of recirculation zones in the hourglass design, which could lead to significant temperatures in the fuel salt and surrounding materials. These first coupled solutions show why these approaches may be necessary for not only predicting the precursor drift effect in fast MSRs but also for reactor design and performance assessments.

scholarly journals THE LiCore POWER PLANT SIMULATOR OF THE MOLTEN SALT FAST REACTOR

2021 ◽  
Vol 247 ◽  
pp. 06030
Author(s):  
A. Laureau ◽  
E. Rosier ◽  
E. Merle ◽  
S. Beils ◽  
O. Bruneau ◽  
...  
Keyword(s):  
Power Plant ◽  
Molten Salt ◽  
Basic Principle ◽  
Kinetics Model ◽  
Simple Point ◽  
Delayed Neutron ◽  
Integrated Simulation ◽  
The Core ◽  
Definition Of ◽  
Point Kinetics

Molten salt reactors as liquid-fuelled reactors are flexible in terms of operation or design choices, but they are very different in terms of design, operation and safety approach compared to solid-fuelled reactors. Such reactors call for a new definition of their operating procedures and safety approach. Dedicated developments and studies have been performed in the frame of the European SAMOFAR project of Horizon2020 and in parallel in France involving CNRS, CORYS and Framatome to develop a system code called LiCore adapted to such reactors, corresponding to a basic-principle power plant simulator. The neutronic model LiCore, at the centre of the simulator, corresponds to an improved point-kinetics model to take into account the specificities of a MSR, notably the circulation of the delayed neutron precursors out of the core. Coupled to a simple piston model for the fuel motion in the core, this code can perform calculations faster than real time to simulate the behaviour of the fuel circuit. Transient calculations performed with LiCore are presented, together with comparisons first to a simple point-kinetics model and then to 3D calculations with the TFM-OpenFOAM coupled code. Finally, the LiCore code has recently been integrated in the ALICES platform, the integrated simulation toolset designed by CORYS for the development, maintenance and operation of major simulator such as power plant simulators.

COUPLE, A Time-Dependent Coupled Neutronics and Thermal-Hydraulics Code, and its Application to MSFR

2014 ◽  
Author(s):  
Dalin Zhang ◽  
Zhi-Gang Zhai ◽  
Andrei Rineiski ◽  
Zhangpeng Guo ◽  
Chenglong Wang ◽  
...  
Keyword(s):  
Molten Salt ◽  
Liquid Fuel ◽  
Time Dependent ◽  
Molten Salt Reactor ◽  
Thermal Hydraulics ◽  
Delayed Neutron ◽  
Fuel Flow ◽  
Reactor Optimization ◽  
Fuel Characteristics ◽  
Significant Attention

Molten salt reactor (MSR) using liquid fuel is one of the Generation-IV candidate reactors. Its liquid fuel characteristics are fundamentally different from those of the conventional solid-fuel reactors, especially the much stronger neutronics and thermal hydraulics coupling is drawing significant attention. In this study, the fundamental thermal hydraulic model, neutronic model, and some auxiliary models were established for the liquid-fuel reactors, and a time-dependent coupled neutronics and thermal hydraulics code named COUPLE was developed to solve the mathematic models by the numerical method. After the code was verified, it was applied to the molten salt fast reactor (MSFR) to perform the steady state calculation. The distributions of the neutron fluxes, delayed neutron precursors, velocity, and temperature were obtained and presented. The results show that the liquid fuel flow affects the delayed neutron precursors significantly, while slightly influences the neutron fluxes. The flow in the MSFR core generates a vortex near the fertile tank, which leads to the maximal temperature about 1100 K at the centre of the vortex. The results can provide some useful information for the reactor optimization.

Microvessel mean transit time and blood flow velocity of sulfhemoglobin-RBC

1980 ◽  
Vol 238 (5) ◽  
pp. H745-H749 ◽  
Author(s):  
C. H. Baker ◽  
E. T. Sutton ◽  
D. L. Davis
Keyword(s):  
Blood Flow ◽  
Red Blood Cells ◽  
Flow Velocity ◽  
Transit Time ◽  
Blood Flow Velocity ◽  
Blood Cells ◽  
Optical Measurement ◽  
Mean Transit Time ◽  
The Mean ◽  
Concentration Curves

An indicator dilution technique is described for obtaining time-concentration curves subsequent to bolus injections of sulfhemoglobin red blood cells (SH-RBC), which have a deep greenish-brown color (absorption peak 620 nm vs. 542 and 564 nm for normal red cells). The series- and parallel-coupled microvessels of cat mesentery were studied. This is accomplished by means of video microscopy with a two-window intensity-sensitive video sampler system. The relationship between SH-RBC concentration in blood and optical measurement is linear. Blood flow velocities were calculated from the difference in mean transit times between two points along a vessel. When this technique is used in association with the previously reported method for determining time-concentration curves for the plasma indicator FITC-dextran the mean transit time (t) for red blood cells was less than for plasma in arterioles. The reproducibility of t and flow velocity for both SH-RBC and FITC-dextran from successive injections were reported. The mean transit time ratio of arteriolar SH-RBC to FITC-dextran averages 0.89. Blood flow velocity calculated from SH-RBC is greater than that calculated from FITC-dextran in these same arterioles. The ratio of the velocities averages 1.29.

The Disposal of Tailed Peak in the Time Spectrum of Pulsed Neutron Oxygen Activation Logging

2013 ◽  
Vol 318 ◽  
pp. 567-571
Author(s):  
Wu Dong ◽  
Hai Min Guo ◽  
Xin Qi
Keyword(s):  
Flow Velocity ◽  
Water Flow ◽  
Transit Time ◽  
Gaussian Function ◽  
Oxygen Activation ◽  
Time Spectrum ◽  
Flow Time ◽  
Redundant Data ◽  
Time Flow ◽  
Pulsed Neutron

In the time spectrum of impulse oxygen activation logging, grave tailed peak occurs when the viscosity of polymer is high and the velocity of the flow is not even. The calculation of transit time, flow velocity as well as flow volume depends on the duration of the oxygen activation peak, so the disposal of the grave tailed peak in the time spectrum of impulse oxygen activation logging is critical. And an effective way to deal with tailed peak will improve the calculation accuracy of transit time and flow velocity. In this paper, the problem of tailed peak is studied, then an algorithm combining Logarithmic transformation and Gaussian function, which is used to deal with tailed peaks in the time spectrum of water flow of pulsed neutron oxygen activation logging, is proposed. Gaussian function is used to describe tailed peaks in the time spectrum of water flow of pulsed neutron oxygen activation logging in the logarithm space. Case analysis shows that the disposal effect of the algorithm is good, effectively eliminating the redundant data of the tailed peak in the flow time spectrum of pulsed neutron oxygen activation logging. The algorithm is a practicable method to deal with the tailed peak in the flow time spectrum of pulsed neutron oxygen activation logging.

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