Liquid-fueled molten salt reactors (MSRs) are usually considered nonclassical reactor types because of the specific nature of the fuel, which is typically constituted by a molten fluoride salt mixture circulating in the primary circuit. The fission material (uranium and/or transuranium elements) is dissolved in the molten salt carrier, which also acts as coolant. Thanks to the potentialities of this liquid fuel, several MSR concepts were investigated at Oak Ridge National Laboratory in the past (see www.energyfromthorium.com/pdf/), and in recent years MSRs have been the subject of renewed interest in the framework of Generation IV nuclear reactors (GIF, 2002, 2014; Serp et al., 2014; IRSN, 2015). These concepts differ mainly by neutron balance (critical or subcritical), neutron spectra (thermal, epithermal, or fast), the presence/absence of the graphite matrix as moderator, and the fuel salt chemical composition. The physics of circulating nuclear fuels involves a strong coupling between neutronics and thermo-hydrodynamics, which would require in general the adoption of a multiphysics modeling approach (e.g., see, Luzzi et al., 2012b, and also Chapter 25, Research activities of this book). However, in this chapter, analyses are performed assuming that the neutronic term is decoupled from fluid dynamics, and appears like a heat source within the fuel/coolant molten salt. The aim is to investigate only the thermo-hydrodynamic behavior. Reference is made to a simple axial-symmetric cylindrical geometry representative of a typical graphite moderated MSR power channel, taking into account the thermodynamic and transport properties of the molten salt as well as its local flow conditions and heat transfer. Even if this assumption simplifies the equations to be solved, the thermohydrodynamic behavior of the molten salt remains complex. In this context, a preliminary analytic approach (Di Marcello et al., 2008) to evaluate the temperature radial profile in both fuel and graphite is reported in Sections 6.2 and 6.3, which are intended to offer the reader a useful validation framework for testing more sophisticated computer codes, in view of their adoption for more realistic and complex 3-D geometry analyses. The circulating "already molten" fuel offers positive peculiarities to be exploited in the safety approach as well as in the fuel cycle of liquid-fueled MSRs (LeBlanc, 2010; Luzzi et al., 2012a; Krepel et al., 2014). For instance, the fluid nature of the fuel means that the reactor core meltdown is an irrelevant instance. Moreover, the reactor has almost no excess of nuclear reactivity, which reduces the risk of accidental reactivity insertion. On the contrary, the decay heat produced by the liquid fuel dissolved into the molten salt and distributed along a closed loop may impair the natural circulation features, leading to an undesired behavior of the reactor. Actually, natural circulation in the presence of internal heat generation (IHG) is characterized by a particular dynamics that needs to be carefully studied. In this context, Section 6.4 presents a preliminary investigation of IHG effects on natural circulation, with reference to the stability maps of single-phase rectangular loops.

Thermal hydraulics of liquid-fueled MSRs

LUZZI, LELIO;CAMMI, ANTONIO;DI MARCELLO, VALENTINO;PINI, ALESSANDRO
2017

Abstract

Liquid-fueled molten salt reactors (MSRs) are usually considered nonclassical reactor types because of the specific nature of the fuel, which is typically constituted by a molten fluoride salt mixture circulating in the primary circuit. The fission material (uranium and/or transuranium elements) is dissolved in the molten salt carrier, which also acts as coolant. Thanks to the potentialities of this liquid fuel, several MSR concepts were investigated at Oak Ridge National Laboratory in the past (see www.energyfromthorium.com/pdf/), and in recent years MSRs have been the subject of renewed interest in the framework of Generation IV nuclear reactors (GIF, 2002, 2014; Serp et al., 2014; IRSN, 2015). These concepts differ mainly by neutron balance (critical or subcritical), neutron spectra (thermal, epithermal, or fast), the presence/absence of the graphite matrix as moderator, and the fuel salt chemical composition. The physics of circulating nuclear fuels involves a strong coupling between neutronics and thermo-hydrodynamics, which would require in general the adoption of a multiphysics modeling approach (e.g., see, Luzzi et al., 2012b, and also Chapter 25, Research activities of this book). However, in this chapter, analyses are performed assuming that the neutronic term is decoupled from fluid dynamics, and appears like a heat source within the fuel/coolant molten salt. The aim is to investigate only the thermo-hydrodynamic behavior. Reference is made to a simple axial-symmetric cylindrical geometry representative of a typical graphite moderated MSR power channel, taking into account the thermodynamic and transport properties of the molten salt as well as its local flow conditions and heat transfer. Even if this assumption simplifies the equations to be solved, the thermohydrodynamic behavior of the molten salt remains complex. In this context, a preliminary analytic approach (Di Marcello et al., 2008) to evaluate the temperature radial profile in both fuel and graphite is reported in Sections 6.2 and 6.3, which are intended to offer the reader a useful validation framework for testing more sophisticated computer codes, in view of their adoption for more realistic and complex 3-D geometry analyses. The circulating "already molten" fuel offers positive peculiarities to be exploited in the safety approach as well as in the fuel cycle of liquid-fueled MSRs (LeBlanc, 2010; Luzzi et al., 2012a; Krepel et al., 2014). For instance, the fluid nature of the fuel means that the reactor core meltdown is an irrelevant instance. Moreover, the reactor has almost no excess of nuclear reactivity, which reduces the risk of accidental reactivity insertion. On the contrary, the decay heat produced by the liquid fuel dissolved into the molten salt and distributed along a closed loop may impair the natural circulation features, leading to an undesired behavior of the reactor. Actually, natural circulation in the presence of internal heat generation (IHG) is characterized by a particular dynamics that needs to be carefully studied. In this context, Section 6.4 presents a preliminary investigation of IHG effects on natural circulation, with reference to the stability maps of single-phase rectangular loops.
Molten Salt Reactors and Thorium Energy
978-0-08-101126-3
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Utilizza questo identificativo per citare o creare un link a questo documento: http://hdl.handle.net/11311/1044110
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