Conference Agenda

Molten salt reactors (MSRs) are a promising reactor type, offering excellent safety and economic benefits due to the stable properties of molten salt coolant at high temperatures. However, the relatively high freezing point of molten inorganic salts poses a risk of coolant solidification, potentially blocking pipelines when flowing through colder sections. This study investigates the process of molten salt filling in cold pipes, including solidification-melting behaviors, and analyzes the pressure drop variation to estimate pipe clogging caused by freezing. Using the Volume of Fluid (VOF) method and the enthalpy-porosity model, the commercial CFD code ANSYS Fluent is employed to numerically simulate the filling process. Results reveal that a solidification layer forms near the cold wall, while the high-temperature incoming flow melts the layer first at the inlet, with the layer thickness decreasing along the flow direction. Analysis shows that the mushy zone constant (A_mush) significantly impacts flow pressure loss, particularly for lower inlet temperatures in finite-length pipes. Higher values of ​ accelerate pressure loss growth, though this increase remains below the maximum upstream pressure head. Comparison with experimental data from Zhang W. demonstrates that the estimated blockage penetration distances for HTS at A_mush=5×10⁴ exhibit an error within 30%. This highlights the critical importance of selecting an appropriate mushy zone constant to accurately predict solidification processes when using the enthalpy-porosity method.

In molten salt reactors (MSRs), liquid fuel offers benefits like high economic efficiency, safety, and low radioactive waste. This fuel, typically a fluoride- or chloride-based salt, contains soluble fissile material in a carrier salt. Compared to water coolant, the working fluids in MSRs have higher melting points and greater corrosivity. Insoluble fission products generated in the core interact with these fluids, which can threaten the integrity of reactor structures such as pumps. Simplifying the primary system is proposed to enhance MSR safety and integrity. This study introduces the passive molten salt fast reactor (PMFR) to simplify the primary system. The PMFR design removes pumps and relies on natural circulation, increasing safety and simplifying reactor design. However, overly simplified core designs can cause imbalanced flow and unexpected heat removal, affecting reactor power. Therefore, stabilizing core flow while minimizing pressure drop is essential. This paper validates the guide structure performance of PMFR using the multi-physics code GeN-Foam, based on OpenFOAM, which models various physics, including neutronics, thermal-hydraulics, and structural thermal-mechanics. Long-term pseudo-steady operation simulations of PMFR demonstrate its feasibility in achieving target power. Results show encouraging performance under normal operating conditions and suggest further improvements to enhance PMFR safety and economics.

The international maritime organization (IMO) has imposed serious restriction on the carbon emission of the commercial shipping industry, which now accounts for nearly 5% of the world amount. Nuclear power can be an important alternative for supplying the large ships with long-durance and near-zero-carbon-emission energy. The Fluoride-salt-cooled High-temperature Reactors adopting the low-operation-pressure fluoride salt as the coolant and the TRISO-particle fuel, show great advantages in the inherent safety, high economics, and reduced difficulty in the licensing. Thus the FHRs can be good reactor concept candidate for the commercial ships. The wind and wave in the ocean bring about oscillation to the flow in the reactor system, which leads to the periodic variation of the heat transfer between the salt and fuel. In further, the safety performance of the shipping-applied FHRs will also be influenced. The Nuclear Reactor Thermal-hydraulic Lab in Shanghai Jiao Tong University has explored the influence of the ocean environment on FHRs, including the core thermal-hydraulics and safety evaluation. The flow regime transition under the pulsation flow is explored, with the transition criteria determined for different pulsation amplitude and period. The system analysis code is developed with the implementation of the additional force models. The system code is validated through the scaled integral effects experimental data. Finally the safety performance under the inclination, heaving and rolling motions is obtained.

In light of the licensing process of advanced reactor designs, a fundamental step to support the safety assessment consists of identifying and quantifying the uncertainties resulting from a lack of extensive practical knowledge and modelling assumptions. The uncertainty characterisation imposes specific requirements for the numerical tools employed to inspect safety-related phenomena. When dealing with Molten Salt Reactors (MSRs), the inherent characteristics of circulating fuel result in the need to perform multidimensional and multiphysics simulations to investigate the steady state and dynamic behaviour of the MSR concept. The multiphysics approach allows to capture the relevant governing phenomena strictly related to the strong coupling between neutronics and thermal-hydraulics. On the other hand, in the context of safety analysis, system codes have proven their suitability to represent the whole plant behaviour, implement submodules devoted to uncertainty quantification, and calibrate models with experimental data. In this work, a computational chain coupling well elaborated system codes and high-fidelity multiphysics tools is developed to manage in the same environment different levels of detail. The modelling framework couples Modelica and OpenFOAM modelling tools thanks to Functional Mock-Up Interfaces, which define a container and an interface to exchange dynamic simulation models. This approach embraces a multidimensional and multiphysics model of the MSR core while preserving a global representation of the plant. The OpenFOAM-Modelica coupling chain is tested on a case study involving a symmetric portion of the Molten Salt Fast Reactor primary loop with a simplified representation of the intermediate salt circuit and Balance of Plant.

The Molten Salt Reactor (MSR) represents a prominent Generation IV design, addressing the urgent need for safer and more sustainable nuclear energy production. This study aims to capture the multi-physics behavior of the Molten Salt Reactor Experiment (MSRE), with a particular focus on thermal-hydraulic and neutronic interactions within the primary loop. Utilizing the GeN-Foam code, we implement detailed Computational Fluid Dynamics (CFD) and heat transfer models to enhance the accuracy of turbulence, drag forces, and porous media characteristics.

Benchmarking against data from Oak Ridge National Laboratory (ORNL) confirms the robustness of this approach, with simulation values closely aligning with recorded ORNL data. For instance, the fuel velocity in the core exhibited a deviation of merely 0.84% from ORNL data, while the pressure at the MSRE core was within 0.964% of the recorded values. Furthermore, temperature measurements at the fuel inlet and outlet demonstrated minimal deviations of 0.039% and 0.008%, respectively. These results provide critical safety insights by elucidating feedback mechanisms that influence neutronics, thermal-hydraulics, and structural integrity.

Significantly, this model, based on the established multi-physics framework of GeN-Foam, can be adapted for other MSR designs through modifications to geometry and input parameters, obviating the need for further code development. The findings from this research offer valuable insights for optimizing MSR designs and safety evaluations, thereby contributing to future regulatory and developmental applications in the field of nuclear technology.

This paper describes implementation of Molten Salt – Concrete Interactions (MSCI) into the System Thermal-Hydraulic code SPECTRA. It consists of two parts:

• The first part discusses the Molten Core – Concrete Interactions (MCCI) model, applicable to the LWR fuel, which had been implemented at an earlier stage.
• The second part discusses the MSCI model, applicable to MSRs.

The summary and conclusions are presented below.

• MCCI. The MCCI model appropriate for LWR fuel was validated based on comparisons with other code (MELCOR) results. Recommendations for future work include:
  • Implementation of reactions such as melting of quartz, decomposition of hematite into magnetite, melting of magnetite.
  • Implementation of reduction of SiO₂ and Fe₂O₃ by Zr.
  • Validation against available experimental data.
• MSCI. Due to lack of suitable data, a validation of the model by comparison with measured data or even code-to-code comparisons is currently not possible. Two test runs were performed as a verification and qualitative validation of the model.
  • Verification is performed by comparing MSCI results with MCCI results in a hypothetical case where the fuel-salt mixture properties are the same as the corium (UO₂-ZrO₂) properties.
  • Qualitative verification is performed by using realistic (as far as possible) data and checking if the qualitative behavior is in agreement with what is expected.