Transient Analysis of Small Molten Salt Reactor : interactions between fission reaction and fuel salt flow in the case of blockage accident

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.

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

Volume 3: Thermal Hydraulics; Instrumentation and Controls ◽

10.1115/icone16-48096 ◽

2008 ◽

Cited By ~ 1

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.

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Natural Convection With Non-Uniform Heat Generation in Molten Salt Reactor

Volume 4: Codes, Standards, Licensing and Regulatory Issues; Student Paper Competition ◽

10.1115/icone17-75180 ◽

2009 ◽

Cited By ~ 1

Author(s):

Libo Qian ◽

Suizheng Qiu ◽

Dalin Zhang ◽

Guanghui Su

Keyword(s):

Natural Convection ◽

Rayleigh Number ◽

Molten Salt ◽

Heat Generation ◽

Dimensionless Temperature ◽

Molten Salt Reactor ◽

Closed Cavity ◽

Dimensionless Velocity ◽

Fuel Salt ◽

Salt Flow

The Molten Salt Reactor (MSR) is one of the six Generation IV systems capable of breeding and transmutation of actinides and long-lived fission products, which uses the liquid molten salt as the fuel solvent, coolant and heat generation simultaneously. The MSR neutronics, such as the distribution of the delay neutron precursors (DNP), is significantly influenced by the fluid flow, which is quite different from the conventional reactors. Therefore, it is very important to do some research on MSR, especially in accident conditions. The present paper studies the natural convection through which the heat generated by the fuel is removed out of the core region (simply a square cavity in this paper). The neutronic theoretical model is founded based on the conservation law, which consists of two-group neutron diffusion equation for the fast and thermal neutron fluxes and that for one-group DNP, in which the convection terms are included to reflect the fuel salt flow. The SIMPLER numerical method was used to calculate the natural convection heat transfer to the molten salt inside a closed cavity for which the boundary temperature was spatially uniform. The equations were discretized by finite volume method based on collocated grids, in which QUICK defect correction was adopted for the convection terms and the central difference was for the diffusion terms. The discretization equations were calculated by ADI (Alternative Direction Implicit) with block-correction technique. The distributions of the dimensionless temperature, the dimensionless velocity, the fluxes and the DNP in the cavity were obtained. The calculated results showed that: a) the distribution of the DNP was correlated both with that of the fluxes and with the fuel salt flow and when Rayleigh number increased, the latter one was of much more importance; b) the distribution of the local Nusselt number varied with different Rayleigh numbers; c) the distribution of the dimensionless velocity and the dimensionless temperature were also closely related to Rayleigh number; d) the maximum dimensionless temperature decreased as Rayleigh number increases.

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

12th International Conference on Nuclear Engineering, Volume 1 ◽

10.1115/icone12-49077 ◽

2004 ◽

Cited By ~ 3

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.

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Molten Salt Reactor Characteristics Involving Fuel Salt Flow and Temperature Distribution in the Steady State

Transactions of the Atomic Energy Society of Japan ◽

10.3327/taesj2002.2.400 ◽

2003 ◽

Vol 2 (4) ◽

pp. 400-407

Author(s):

Koshi MITACHI ◽

Takahisa YAMAMOTO ◽

Takashi SUZUKI ◽

Yoshihide OKUMURA

Keyword(s):

Steady State ◽

Temperature Distribution ◽

Molten Salt ◽

Molten Salt Reactor ◽

Fuel Salt ◽

Salt Flow

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

Radiochimica Acta ◽

10.1515/ract-2021-1011 ◽

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.

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

Volume 2: Plant Systems, Structures, and Components; Safety and Security; Next Generation Systems; Heat Exchangers and Cooling Systems ◽

10.1115/icone20-power2012-54384 ◽

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.

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Steady Thermal Hydraulics Analysis for a Graphite-Moderated Channel Type Molten Salt Reactor

Volume 4: Structural Integrity; Next Generation Systems; Safety and Security; Low Level Waste Management and Decommissioning; Near Term Deployment: Plant Designs, Licensing, Construction, Workforce and Public Acceptance ◽

10.1115/icone16-48944 ◽

2008 ◽

Author(s):

Dalin Zhang ◽

Suizheng Qiu

Keyword(s):

Molten Salt ◽

Flow Distribution ◽

Axial Direction ◽

Theoretical Models ◽

Outer Wall ◽

Molten Salt Reactor ◽

Thermal Hydraulics ◽

The Core ◽

Fuel Salt ◽

Channel Type

The Molten Salt Reactor (MSR) is one of the six GENIV systems capable of breading and burning. In this paper, a graphite-moderated channel type MSR was selected for conceptual research. For this MSR, a ternary system of 0.15LiF-0.58NaF-0.27BeF2 was proposed as the reactor fuel solvent, coolant and also moderator simultaneously with ca.1 mol% UF4 dissolving in it, which circulates through the whole primary loop accompanying fission reaction only in the core. 169 hexagonal graphite elements, each with a central fuel channel, are arranged in the core symmetrically by 30° angles. The theoretical models of the thermal hydraulics under steady condition are conducted in one-twelfth of the core and calculated by the numerical method. The DRAGON code is adopted to calculate the axial and radial power factors. The flow and heat transfer models in the fuel salt and graphite are founded basing on the fundamental mass, momentum and energy equations. The calculated results show the detailed mass flow distribution in the core; and the temperature of the fuel salt, inner and outer wall in the calculated elements along the axial direction are also obtained.

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