Selected Articles
Similarity Analysis of High-Prandtl Surrogate Fluids for Thermal-Hydraulic Studies of Molten Salt Reactors
Annals of Nuclear Energy Hossam H. Abdellatif, David Arcilesi, Richard Christensen, Arsen Iskhakov
2025-08-05
Molten salt reactors (MSRs) require high-temperature testing of corrosive fluids like FLiBe and FLiNaK, complicating integral and separate effect experiments. To overcome these limitations, this study employs a similarity-based methodology, grounded in the hierarchical two-tier scaling (H2TS) framework to investigate the performance of Therminol-66, DOWTHERM A, and DOWTHERM RP as low-temperature, high-Prandtl-number surrogates for fluoride salts. We first match each simulant’s Prandtl number to its corresponding salt at the reactor’s mean operating temperature, then derive all relevant non-dimensional groups under forced and natural circulation. Applying this to three MSR designs, KP-FHR, FuSTAR, and FUJI-233Um, yields compact, low-power test loops with zero distortion in all primary groups and negligible Biot-number error. Nusselt-number predictions using standard correlations show that heat transfer is replicated to within 1% over the target Re-Pr range. These results establish a practical, safe, and scalable pathway for high-fidelity thermal-hydraulic experiments supporting future MSR developments.
Flow Instabilities in boiling channels and their suppression methodologies—A review
Nuclear Engineering and Design Hossam H Abdellatif, Walter Ambrosini, David Arcilesi, Palash K Bhowmik, Piyush Sabharwall
2024-03-18
Small modular reactors (SMRs) are gaining significant attention as a promising solution for clean and sustainable nuclear-power generation. However, the operation of SMRs is subject to various challenges, including two-phase flow instabilities. Flow instability has the potential to trigger flow-induced vibration and cyclic fluctuations in local thermal stress. These instabilities frequently manifest because of the complex interplay among a multitude of factors, encompassing thermal-hydraulic conditions, the geometric configuration of the steam generator, and operational parameters. These conditions could subsequently lead to premature critical heat flux, equipment malfunctions, and other safety concerns. The endeavor to address steam-generator flow instabilities is of utmost importance in augmenting the sustainability and efficiency of contemporary energy production. This study offers a comprehensive review of instabilities in two-phase flow, with a particular focus on the influential factors impacting the stability of flow boiling. Furthermore, it delves into the processes of identifying, characterizing, and ameliorating these instabilities, emphasizing pivotal findings, methodologies employed, and avenues for prospective research. The primary parameters of concern encompass the efficient transfer of thermal energy, the optimization of mass-flow rates, and the establishment of favorable boundary conditions, all in the context of steam generator design to alleviate instability for water-cooled SMRs. These discernments bear substantial ramifications for enhancing reactor performance and ensuring operational safety.
Preliminary experimental validation of single-phase natural circulation loop based on RELAP5-3D code: Part I
Journal of Nuclear Engineering Hossam H Abdellatif, Joshua Young, David Arcilesi, Richard Christensen
2025-09-18
The molten salt reactor (MSR) is a prominent Generation IV nuclear reactor concept that offers substantial advantages over conventional solid-fueled systems, including enhanced fuel utilization, inherent passive safety features, and significant reductions in long-lived radioactive waste. Central to its safety strategy is a reliance on natural circulation (NC) mechanisms, which eliminate the need for active pumping systems and enhance system reliability during normal and off-normal conditions. However, the challenges associated with molten salts, such as their high melting points, corrosivity, and material compatibility issues, render experimental investigations inherently complex and demanding. Therefore, the use of high-Pr-number surrogate fluids represents a practical alternative for studying molten salt behavior under safer and more accessible experimental conditions. In this study, a single-phase natural circulation loop setup at the University of Idaho’s Thermal–Hydraulics Laboratory was employed to investigate NC behavior under various operating conditions. The RELAP5-3D code was initially validated against water-based experiments before employing Therminol-66, a high-Prandtl-number surrogate for molten salts, in the natural circulation loop for the first time. The RELAP5-3D results demonstrated good agreement with both steady-state and transient experimental results, thereby confirming the code’s ability to model NC behavior in a single-phase flow regime. The results also highlighted certain experimental limitations that should be addressed to enhance the NC loop’s performance. These include increasing the insulation thickness to reduce heat losses, incorporating a dedicated mass flow measurement device for improved accuracy, and replacing the current heater with a higher-capacity unit to enable testing at elevated power levels. By identifying and addressing the main causes of these limitations and uncertainties during water-based experiments, targeted improvements can be implemented in both the RELAP5 model and the experimental setup, thereby ensuring that tests using a surrogate fluid for MSR analyses are conducted with higher accuracy and minimal uncertainty.
Evaluating direct vessel injection accident-event progression of AP1000 and key figures of merit to support the design and development of water-cooled small modular reactors
Nuclear Engineering and Technology
Hossam H Abdellatif, Palash K Bhowmik, David Arcilesi, Piyush Sabharwall
2024-02-14
The passive safety systems (PSSs) within water-cooled reactors are meticulously engineered to function autonomously, requiring no external power source or manual intervention. They depend exclusively on inherent natural forces and the fundamental principles of reactor physics, such as gravity, natural convection, and phase changes, to manage, alleviate, and avert the release of radioactive materials into the environment during accident scenarios like a loss-of-coolant accident (LOCA). PSSs are already integrated into such operating commercial reactors as the Advanced Pressurized Reactor-1000 MWe (AP1000) and the Water-Water Energetic Reactor-1200 MWe (WWER-1200) are adopted in most of the upcoming small modular reactor (SMR) designs. Examples of water-cooled SMR PSSs are the passive emergency core-cooling system (ECCS), passive containment cooling system (PCCS), and passive decay-heat removal system, the designs of which vary based on reactor system-design requirements. However, understanding the accident-event progression and phases of a LOCA is pivotal for adopting a specific PSS for a new SMR design. This study covers the accident-event progression for direct vessel injection (DVI) small-break loss-of-coolant accident (SB-LOCA), associated physics phenomena, knowledge gaps, and important figures of merit (FOMs) that may need to be evaluated and assessed to validate thermal-hydraulics models with an available experimental dataset to support new SMR design and development.
Numerical Investigation of Air Natural Convection in the AP1000 Passive Containment Cooling System Following LBLOCA Using ANSYS FLUENT
Nuclear Science and Engineering
Hossam H Abdellatif, David Arcilesi
2024-08-01
The innovative design of the AP1000 power plant has various layers of passive safety systems aiming to enhance reactor safety during normal and transient conditions. The passive containment cooling system (PCCS) is a safety-related system capable of removing heat from the steel containment vessel (SCV) to the atmosphere and preventing the containment from exceeding the design pressure and temperature following a postulated design-basis accident. The PCCS heat removal mechanisms include condensation on the internal SCV surface, heat conduction, natural convection, evaporation of water film, and radiative heat transfer. In two basic postulated scenarios, the reactor decay heat can ultimately be removed from the SCV only by air natural convection. The first scenario occurs 72 h following a large-break loss-of-coolant accident (LBLOCA) when the passive containment cooling water storage tank becomes unavailable. The second scenario occurs following a postulated loss of shutdown decay heat removal event. Hence, investigating the thermal-hydraulic behavior of the containment under transient conditions is essential to ensure its safety and integrity. In this study, a simplified three-dimensional model using ANSYS FLUENT is developed to investigate the cooling capability of air natural convection outside the SCV during a LBLOCA event. Because of the lack of experimental data, code-to-code validation was performed using the actual results of AP1000 alongside other research findings. The results show good agreement with available data, which can be used for future research.
