Molten Salt Coolant Research Articles

Coupled computational fluid dynamics and computational thermodynamics simulations for fission product retention and release: A molten salt fast reactor application

This study presents a computational capability for fission product retention and release in two-phase, multi-species systems representing Molten Salt Reactors (MSR) with coupled thermal-hydraulics and fuel coolant chemical behaviours. This is demonstrated through four simulated cases centred on the proposed Molten Salt Fast Reactor (MSFR). This is achieved by two-way coupling the Computational Fluid Dynamics (CFD) code OpenFOAM and the Computational Thermodynamics (CT) code Thermochimica, using the Joint Research Centre Molten Salt Database (JRCMSD). Local chemical equilibrium is assumed, implying that chemical kinetics are predominantly governed by mass transport. Four simulations address normal operating conditions, exploring: (i) dilution of fission products injected within the molten salt coolant, (ii) molten salt coolant evaporation rate, (iii) release of radioactive gaseous species, (iv) shifts in the UF4/UF3 ratio, and (v) comparison of vapour pressures of gaseous species. The influence of temperature-dependent viscosity on retaining fission products, compared to consistent values, is also discussed. The feasibility of integrating CFD with Thermochimica showed promising results, broadening insights into multiphysics systems and setting the stage for its application in more intricate scenarios.

Thermoelectrics for nuclear fusion reactors: opportunities and challenges

In this review, we discuss the promising applications and practical considerations of thermoelectrics to harvest the unutilized thermal gradient between the plasma-facing surfaces and the molten salt coolant loop in tokamak fusion reactors.

Core design study of the Wielenga Innovation Static Salt Reactor (WISSR)

This paper presents the design features and preliminary design analysis results of the Wielenga Innovation Static Salt Reactor (WISSR). The WISSR incorporates features that make it both flexible and inherently safe. It is based on innovative technology that controls a nuclear reactor by moving molten salt fuel into or out of the core. The reactor is a low-pressure, fast spectrum transuranic (TRU) burner reactor. Inherent shutdown is achieved by a large negative reactivity feedback of the liquid fuel and by the expansion of fuel out of the core. The core is made of concentric, thin annular fuel chambers containing molten fuel salt. A molten salt coolant passes between the concentric fuel chambers to cool the core. The core has both fixed and variable volume fuel chambers. Pressure, applied by helium gas to fuel reservoirs below the core, pushes fuel out of a reservoir and up into a set of variable volume chambers. A control system monitors the density and temperature of the fuel throughout the core. Using NaCl-(TRU,U)Cl3 fuel and NaCl–KCl–MgCl2 coolant, a road-transportable compact WISSR core design was developed at a power level of 1250 MWt. Preliminary neutronics and thermal-hydraulics analyses demonstrate the technical feasibility of WISSR.

A novel nonlinear radiative heat exchanger for molten-salt applications

One of the main difficulties associated with using molten salts as heat transfer fluids (HTFs) in practical applications such as in the solar and nuclear energy industries is the possibility of solidifying molten salts. Since the salts proposed for Generation-IV nuclear reactors (MSRs) have higher melting points, this is especially true for those reactors. The consequences of accidental freezing within the direct reactor auxiliary cooling system (DRACS) of MSRs could be catastrophic. DRACS is a passive safety system designed to remove decay heat from the reactor core during an emergency. The present study examines, as a case study, the freezing problem in the most vulnerable component of DRACS, namely the natural draft heat exchanger (NDHX). For this application, a unique and innovative heat exchanger design termed radiative heat exchanger (RHX) is proposed, with nonlinear heat transfer characteristics. Specifically, by suppressing convection and exploiting thermal radiation as the primary heat transfer mechanism, the RHX concept reduces the heat rejection/removal rate passively to allow safe operation when the molten salt coolant temperature drops and is near its melting point, while maintaining a high capacity for decay heat extraction in case the coolant temperature rises during an accident and the reactor becomes overheated; this moderation of the heat removal rate is accomplished without the requirement for external (manual or automatic) control. A thermohydraulic model is developed and applied to evaluate the proposed RHX with the existing conventional NDHX. Compared to the conventional design, the RHX has the potential to reduce heat removal by approximately 30 % at low temperatures and near-solidification conditions. In the case of the prototypical 20 MW reactor with a decay rate of 0.2 MW, RHX extends the freezing time of molten salt by three times, to approximately 18 h. During high-temperature events, such as the initial stages of a reactor scram, the RHX can improve decay heat extraction by 17 %. Other applications of molten salts with similar challenges can also be addressed using the proposed design approach.

Optimal Radial Build and Transmutation Properties of a Fusion-Based Transmutation Reactor with Molten Salt Coolants

The optimal shape of a fusion-based transmutation reactor with a molten salt coolant was determined by plasma physics, technology, and neutronic requirements. System parameters such as neutron multiplication, power density, shielding, and tritium breeding, were calculated in a self-consistent manner by coupling neutron transport analysis with conventional tokamak systems analysis. The plasma physics and engineering levels were similar to those used in the International Thermonuclear Experimental Reactor. The influence of aspect ratio of the tokamak and fusion power on the radial build, and the transmutation properties associated with two molten salt options, FLiBe and FliNaBe, were investigated. Being compared with a transmutation reactor with a small aspect ratio, a transmutation reactor with large aspect ratio was smaller in size and had a larger maximum fusion power. This type of reactor also revealed increased tritium-breeding capability and a smaller initial transuranic (TRU) inventory with a slightly lower burn-up rate. The burn-up rate for molten salt using either FLiBe or FLiNaBe was similar, but the initial TRU inventory and the tritium-breeding capability were smaller with FLiNaBe compared with FLiBe.

Experimental research on convective heat transfer characteristics of molten salt in a pebble bed channel with internal heat source

Convective heat transfer between the molten salt coolant and the fuel pebbles in the reactor core is one of the key thermal hydraulic phenomena for Thorium Molten Salt Reactors. In this study, a pebble bed experiment facility using induction heating to provide internal heat source is designed. The heat transfer characteristics of molten salt in the pebble bed channels are studied on the Heat Transfer Salt (HTS) loop. The effects of inlet temperature, molten salt flow rate and induction heating power are investigated on convective heat transfer characteristics. A new correlation is proposed for predicting the convective heat transfer coefficients of molten salt with Prandtl number at 11.27–14.51 and Reynolds number at 2800–6600. The results will provide an important reference for the core design of thorium based molten salt reactor.

Hydrogen and its detection in fusion and fission nuclear materials – a review

Fusion and fission reactions are profoundly dependent on hydrogen for sustained reactions. Fusion is fueled by the isotopes of hydrogen, and predominantly hydrogen-based moderators slow fission neutrons to propagate chain reactions. Intentional tritium production in fusion and concomitant tritium production in fission reactors introduce challenges. The technology to efficiently extract, harvest, and quantify tritium in and from advanced fission molten salt coolants and fusion molten tritium breeder materials will require more pronounced research and development. Measuring and quantifying hydrogen is necessary in all areas of nuclear materials. Although many characterization techniques cannot directly detect hydrogen, numerous techniques provide the necessary information to understand the behavior of hydrogen in nuclear materials.

Identification of surrogate fluids for molten salt coolants used in energy systems applications including concentrated solar and nuclear power plants

Identification of surrogate fluids for molten salt coolants used in energy systems applications including concentrated solar and nuclear power plants

Analysis of the Power and Temperature distribution in molten salt reactors with TRACE. Application to the MSRE

In this study, building on the author’s previous experience in the simulation of the MSRE system, a more detailed power deposition model was considered. We decided to take into account the power generated into the graphite moderator and the decay heat generated inside and outside the reactor core. The work was done using a modified by us version of the system code TRACE capable of simulating molten salt coolants and the transport of neutron precursors with the coolant, and the Monte-Carlo based Serpent code to generate the appropriate core neutronic data used in the point kinetics models of TRACE. The study shows that the temperature field differs significantly taking into account these power sources. The developed reactor model was also used to simulate the transient behaviour of the MSRE. Comparable results for transient behaviour reported in the original MSRE reports by ORNL support the correctness of the model. In the future, it will be used for the other molten salt reactors, for which interests of science and industry is growing.

THE UK NATIONAL THERMAL-HYDRAULICS FACILITY: MOTIVATIONS, DESIGN AND PLANNING STATUS

A UK National Thermal-Hydraulics Facility (NTHF) dedicated to supporting new reactor and other relevant business is being developed, one of the purposes being to deliver on the government’s carbon emission reduction commitments. The facility site is foreseen to be at Menai Science Park on the isle of Anglesey in North Wales, a region expected to see significant low carbon energy deployment in coming years. The UK NTHF is envisioned to cater for the needs of emerging nuclear in the UK – but also to serve as a hardware platform for international thermal-hydraulics research collaboration. Plans are to construct a platform capable of maintaining several test loops including support for the UK’s on-going, conventional nuclear new build programme as well as Gen-IV systems and associated materials like molten salt and liquid metal coolant media. Motivations are given for NTHF expected capabilities and requirements, which form the basis for its current design and planning state.

RESIDUAL HEAT ESTIMATION OF NUCLEAR FUEL IN ENERGY WELL REACTOR

Energy Well (EW) is a concept of a small modular reactor, designed as a source of energy for places with low infrastructure. Its design combines TRISO fuel particles, graphite moderator and molten salt coolant. One of key features is that the active core is small and therefore easy to transport. Residual heat estimation of EW fuel assembly is conducted using two different computation codes, SCALE and SERPENT. The obtained residual heat power is then used as an input in TRACE model of the EW to obtain the development of residual temperature in the coolant in the core inlet and outlet and also maximum temperature on the outer side of a fuel slab. The resulting temperatures are below the operational temperature.

Numerical assessment of packed-bed heat transfer correlations for molten salt

The Fluoride salt-cooled High-temperature Reactor (FHR) is a subset of next-generation Nuclear Reactors that are under active development. The proposed pebble-bed FHR designs use a high Prandtl number molten salt coolant, 2LiF–BeF2, with spherical fuel elements. In order to determine convective heat transfer coefficients, packed-bed correlations available in the literature are used. However, past correlations were developed for either gas or water as the coolant, which have a low Prandtl number (NPr≈1). This study assesses the applicability of such correlations using numerical simulations. A range of Reynolds numbers (102⩽NRe≤2.25·103) is evaluated that encompass the proposed FHR designs (NPr≈16).The study finds that most conventional correlations miss the qualitative and quantitative trend predicted by both kε and LES turbulence models. The Meng correlation (Meng et al., 2012) was found to provide a good agreement with numerical results. Both turbulence models have good agreement up to NRe<103 for surface-averaged parameters. At higher flowrates, the secondary flows are more prominent for the LES solutions – however, the impact on the convective heat transfer coefficient is minimal.

Study on neutronics design of ordered-pebble-bed fluoride-salt-cooled high-temperature experimental reactor

This paper presents a neutronics design of a 10 MW ordered-pebble-bed fluoride-salt-cooled high-temperature experimental reactor. Through delicate layout, a core with ordered arranged pebble bed can be formed, which can keep core stability and meet the space requirements for thermal hydraulics and neutronics measurements. Overall, objectives of the core include inherent safety and sufficient excess reactivity providing 120 effective full power days for experiments. Considering the requirements above, the reactive control system is designed to consist of 16 control rods distributed in the graphite reflector. Combining the large control rods worth about 18000–20000 pcm, molten salt drain supplementary means (− 6980 to − 3651 pcm) and negative temperature coefficient (− 6.32 to − 3.80 pcm/K) feedback of the whole core, the reactor can realize sufficient shutdown margin and safety under steady state. Besides, some main physical properties, such as reactivity control, neutron spectrum and flux, power density distribution, and reactivity coefficient, have been calculated and analyzed in this study. In addition, some special problems in molten salt coolant are also considered, including 6Li depletion and tritium production.

Application of frequency response methods in separate and integral effects tests for molten salt cooled and fueled reactors

Molten salt cooled and fueled reactors can be distinguished from other reactor types by their low volatility liquid coolant, which remains single-phase under most operating and accident conditions, and by the substantial heat capacity of both the coolant and solid structures. Forced and natural circulation of the molten salt coolant provide the primary heat transport mechanisms under power and shutdown operation, but thermal coupling to solid heat structures has important effects on dynamic system response during transients. These characteristics make frequency response methods particularly suitable to characterizing and predicting system response to transients. Predicting the temperature of coolant boundary structures during transients, including accidents, is also important for assessing whether damage may occur due to thermal stresses or accelerated thermal creep. This paper discusses how frequency response methods may be used in separate effect and integral effect tests, particularly with simulant fluids to measure the thermal inertia and coupling properties of heat structures and to validate transient response models. The methodology is discussed, followed by examples for application. The scaling parameters developed in this paper can be expected to play a major role in the design of integral effect test facilities for FHRs and MSRs.

Numerical simulations of molten salt pebble-bed lattices

The development of the Fluoride-salt-cooled High-temperature Reactor (FHR) is ongoing. The molten salt coolant provides several inherent safety characteristics that traditional water-cooled nuclear reactors lack. The first test reactor to be built in China is a pebble-bed type. Research efforts are currently focused on determining the heat transfer characteristics of such a reactor. In the past, the pebble-bed geometry has been studied for the development of the High Temperature Gas-cooled Reactor (HTGR). The previous work has indicated that the profile of the Nusselt number depends on both the lattice structure and pebble contact.In this work, a new series of simulations is sought to characterize heat transfer in an FHR pebble-bed. Due to the physical properties of molten salt and the FHR design, the Reynolds number is roughly three orders of magnitude lower than the HTGR while the Prandtl number is roughly an order of magnitude higher. Thus, the flow in an FHR lattice is expected to be unlike the HTGR counterpart. Three different lattices (Simple Cubic (SC), Body Centered Cubic (BCC), and Face Centered Cubic (FCC)) are considered.The study found that the different lattice geometries have a significant impact on the heat transfer characteristics. The SC lattice has a simplified flow that minimizes pressure-drop and Nusselt number. The BCC and FCC lattice have a complicated flow structure resulting in an order of magnitude higher pressure-drop and almost double the Nusselt number. It was concluded that heat transfer characteristics in a random pebble-bed would be highly sensitive to the mixing, and thus turbulent heat transfer, along the flow path. It was also found that while an interstitial gap allows heat transfer characteristics to be replicated, the pressure loss is significantly lower, while the computational costs increase. Correlations available in the literature (developed primarily for HTGRs) have a mixed performance in prediction of the Nusselt number. The result suggests that new experimental efforts are necessary to develop correlations that pertain to FHR conditions.

Progress in modeling solidification in molten salt coolants

Progress in modeling solidification in molten salt coolants

Progress in modeling solidification in molten salt coolants

Molten salts have been proposed as heat carrier media in the nuclear and concentrating solar power plants. Due to their high melting temperature, solidification of the salts is expected to occur during routine and accidental scenarios. Furthermore, passive safety systems based on the solidification of these salts are being studied. The following article presents new developments in the modeling of eutectic molten salts by means of a multiphase, multicomponent, phase-field model. Besides, an application of this methodology for the eutectic solidification process of the ternary system LiF–KF–NaF is presented. The model predictions are compared with a newly developed semi-analytical solution for directional eutectic solidification at stable growth rate. A good qualitative agreement is obtained between the two approaches. The results obtained with the phase-field model are then used for calculating the homogenized properties of the solid phase distribution. These properties can then be included in a mixture macroscale model, more suitable for industrial applications.

DEM–CFD simulation of modular PB-FHR core with two-grid method

For designing and optimizing the reactor core of modular pebble-bed fluoride salt-cooled high-temperature reactor (PB-FHR), it is of importance to simulate the coupled fluid and particle flow due to strong coolant–pebble interactions. Computational fluid dynamics and discrete element method (DEM) coupling approach can be used to track particles individually while it requires a fluid cell being greater than the pebble diameter. However, the large size of pebbles makes the fluid grid too coarse to capture the complicated flow pattern. To solve this problem, a two-grid approach is proposed to calculate interphase momentum transfer between pebbles and coolant without the constraint on the shape and size of fluid meshes. The solid velocity, fluid velocity, fluid pressure and void fraction are mapped between hexahedral coarse particle grid and fine fluid grid. Then the total interphase force can be calculated independently to speed up computation. To evaluate suitability of this two-grid approach, the pressure drop and minimum fluidization velocity of a fluidized bed were predicted, and movements of the pebbles in complex flow field were studied experimentally and numerically. The spouting fluid through a central inlet pipe of a scaled visible PB-FHR core facility was set up to provide the complex flow field. Water was chosen as liquid to simulate the molten salt coolant, and polypropylene balls were used to simulate the pebble fuels. Results show that the pebble flow pattern captured from experiment agrees well with the simulation from two-grid approach, hence the applicability of the two-grid approach for the later PB-FHR core design.

Development of a GPU-accelerated 3D neutron dynamics code for PB-FHR

Pebble bed fluoride-salt cooled high temperature reactor (PB-FHR) is a kind of novel nuclear energy systems, combining advanced techniques such as coated particle fuel and molten-salt coolant to achieve better safety and economy performance. The compact core of a PB-FHR is geometrically complicated due to the randomly-packed pebble bed with fuel motion in a complex core chamber, which yields a highly detailed core-modeling for accurate neutron dynamics analysis. Three-dimensional fine mesh finite volume method is expected to achieve better accuracy because of its strong geometry adaptability in detailed core-modeling, but is much more time-consuming compared to other methods such as coarse mesh nodal method. A GPU-accelerated 3D fine mesh neutron dynamics code (GAND) is developed, using conjugate gradient method (CG) to solve the time-dependent multi-group neutron diffusion equations in r-z-θ coordinates. The GAND code is verified by a cylindrical reactor benchmark with good agreement, and is preliminarily applied to a PB-FHR core in both static and transient analysis. Speed-up ratio and other performance of GAND code using different equation preconditioners is studied, a best speed-up ratio of 21.65 has been achieved using Neumann polynomials-preconditioned CG.

High temperature reactor technology development in India

High Temperature Reactor technology development programme was initiated in India with an aim to provide high temperature process heat for nuclear hydrogen production by splitting water. As high efficiency hydrogen production needs process heat at temperatures around 1123 K, a challenging technology development goal for the high temperature reactors was set to achieve coolant temperature of 1273 K. Currently development is in progress for a Compact High Temperature Reactor (CHTR), and a 600 MWth Innovative High Temperature Reactor (IHTR). Current design version of CHTR has 235U based TRISO (TRistructural-ISOtropic) coated particle fuel, Beryllium oxide (BeO) as moderator, graphite as reflector, and lead-bismuth eutectic (LBE) as the coolant. The design incorporates many passive safety features for reactor heat removal. Current design version of IHTR is based on pebble bed fuel configuration with molten salt as coolant. For both the reactors, reactor heat is removed passively by natural circulation of the coolant. Technology development for these reactors include development of TRISO coated particle fuel, lead-bismuth eutectic and molten salt coolant technologies, BeO and graphite, oxidation resistant coatings, high creep strength alloys compatible to these coolants, high temperature instrumentation for these coolants, as well as high efficiency hydrogen production and electricity generation technologies.