Understanding of fission products evaporation and retention behaviour in liquid lead is crucial for estimating mechanistic radioactive source term of Lead-cooled Fast Reactors during accident scenarios. An experimental setup was designed and built at the University of New Mexico, to demonstrate the evaporation rate evaluation process of molten lead under oxygen-controlled and oxygen-saturated conditions. Measurements were conducted at a range of liquid lead temperatures between 550 and 700 degrees C, resulting in a highly turbulent argon cover gas of Rayleigh number of approximately 10(8). Measured lead's evaporation rates under oxygen-controlled conditions were consistently higher than predictions from conventional evaporation models, suggesting the presence of localized fog formation near the molten surface. This fog enhances mass transport by steepening the lead vapor concentration gradient within the boundary layer. The evaporation rate increased with increased temperature up to a maximum value of 1.1x10(-6) kg/m(2).sec for oxygen-controlled (similar to 3x10(-6) wt%) and 1.96x10(-7)kg/m(2).sec for oxygen-saturated lead at 700 degrees C. The reduced lead's evaporation rate in the oxygen-saturated lead is attributed to the formation of a surface lead oxide layer, which has lower vapor pressure and inhibits evaporation. The developed experimental platform will support future studies on the evaporation and retention behavior of representative fission product isotopes in molten lead.

The Westinghouse Lead Fast Reactor (LFR) is a 450 MWe class, lead-cooled, fast neutron spectrum, pool-type reactor with passive safety systems. The passive heat removal system (PHRS) is the emergency decay heat removal system of this LFR. A distinctive feature of the PHRS is that it is always on, and does not require any operator intervention, signals of intelligence or moving parts for the actuation and operation, consistent with the IAEA passive safety category B. The system initially removes heat through water boiling and subsequently transitions to (indefinite) air cooling. To assess the performance of the PHRS and provide experimental data for the verification and validation of modeling and simulation tools, the passive heat removal facility (PHRF) has been built at the Ansaldo Nuclear’s site in Wolverhampton, UK, with the support from the United Kingdom Government and international partners. The PHRF is a full-height separate effects test facility that follows the power-to-volume scaling rule to maximize its prototypicality. The test matrix of the first phase of testing included both air cooling tests and transition tests from water cooling to air cooling. A significant amount of experimental data were generated in the first phase of testing, confirming the performance of the PHRS both in air cooling mode and during the transition from water cooling to air cooling.

The nuclear industry has fully embraced the development of accelerated fuel qualification (AFQ) approaches to speed up the assessment and validation of new fuel designs with respect to performance and safety metrics. To support the AFQ approach to shortening the time to develop and qualify new fuel for higher plant performance, Westinghouse utilizes advanced modeling and simulation technologies as part of their integrated and comprehensive AFQ vision through improved fuel performance prediction under various operating conditions and accident scenarios.This paper provides example applications, prioritized in Westinghouse using machine learning technology, for fuel thermal-hydraulic applications with methodologies that are under development for the prediction of critical heat flux for pressurized water reactor (PWR) fuel thermal margin assessment and surrogate model development for crud-induced power shift risk prediction to enhance PWR fuel operation performance.

Lead-cooled fast reactors (LFRs) are part of the so-called Generation IV reactor technologies and use liquid lead as the primary coolant. While lead has several favorable attributes promoting economics, safety, and sustainability, its relatively high freezing point (327 degrees C) requires coolant solidification to be included among the phenomena of interest in the development of the safety case and operational aspects for the nuclear plant. The assessment of current numerical modeling capabilities to predict coolant solidification and tracking of the solidification front propagation is thus essential to the development of LFRs.Motivated by the current development of the Westinghouse LFR, this study presents transient computational fluid dynamics (CFD) simulations of lead solidification within a pool-type geometry cooled externally by forced convection of air using the Siemens STAR-CCM+ code. The geometry is representative of the main vessel of the LEFREEZ test facility built by Westinghouse and its partners as part of Phase 2 of the U.K. Department for Business, Energy and Industrial Strategy Advanced Modular Reactor (AMR) program in the United Kingdom and contains a partially submerged mock-up of a 19-pin fuel assembly. A simplified two-dimensional axisymmetric approach was first applied, and results using four different viscous modeling approaches (laminar, $$k{\rm{ - }}\varepsilon $$k-epsilon, $$k{\rm{ - }}\omega $$k-omega SST, and $$k{\rm{ - }}\varepsilon {\rm{ - }}{v<^>2}{\rm{ - }}f$$k-epsilon-v2-f) predicted the same overall development of the solidification front through the domain, whose shape was influenced by the positioning of the external air inlet beneath the vessel. Simulations of the full three-dimensional model reproduced the solidification front as it moved past the included mock-up fuel bundle and illustrated the relatively complex and nonuniform nature of the external heat transfer. However, the CFD simulations significantly underpredicted the time taken for lead to solidify at each probe location, and consequently, the progression of the solidification front through the bundle was significantly slower. Overall, these results demonstrate that the current melting-solidification modeling capabilities can produce results that are physically consistent with expected behavior, but further work is required to address discrepancies with experimental data.

Lead-cooled fast reactors (LFRs) are Generation IV reactor technologies that use molten lead as the primary coolant. Whilst lead offers advantages for economics, safety, and sustainability, its low Prandlt number and challenging experimental characteristics pose difficulties for thermal-hydraulic modelling and validation. To support LFR development, this study aims to advance modelling capabilities and understanding of the relevant physical phenomena through a series of Computational Fluid Dynamics (CFD) simulations of a Fuel Pin Bundle Simulator (FPBS) that shares design features with the Westinghouse LFR fuel assembly. Three geometrical configurations of the FPBS have been modelled, using the Reynolds-averaged Navier-Stokes (RANS) approach: a bare pin bundle (without spacer grids), a T-junction upstream of the main test section, and the full-length 360 degrees main FPBS test section including spacer grids and instrumentation wires. The sensitivity of the results to modelling choices, including turbulence models and approaches for the turbulent Prandtl number, is explored. The original contributions of this study are in the assessment of different RANS models of the Reynolds stresses, the assessment of different values and functions of the turbulent Prandtl number for the modelling of the turbulent heat fluxes, the exploration of the entry conditions on the flow and thermal development along the fuel bundle and the determinations of the effects of the intrusive instrumentation on the measured quantities. The bare bundle simulations showed only minor sensitivity to the turbulence model and produced friction factors in excellent agreement with existing correlations. Predictions in the upstream T-junction indicated the generation of significant swirl that enters the main test section, but the spacer grid acts as an effective flow straightener. Nusselt number predictions in the main FPBS test section showed good agreement with established correlations for liquid metal rod bundles. Instrumentation wires had only a minor effect on the temperature field and increased the pressure drop by 2.7 %. A sensitivity analysis of the turbulent Prandtl number (Prt) showed that Kay's correlation produced Nusselt numbers that were closest to the empirical correlation of Ushakov et al. (1977), with a mean deviation of 1.1 %. In contrast, a constant Prt= 0.9 resulted in an overprediction of 19 %.

Lead-cooled fast reactors (LFR) are one of the Generation IV reactor technologies that uses a liquid metal as the primary coolant. As part of current development activities of the Westinghouse LFR, we employ CFD models within the RANS framework to conduct pre-test modelling of an experimental Fuel Pin Bundle Simulator (FPBS) that shares design features with the fuel assembly from the Westinghouse LFR. The FPBS comprises 19 electrically heated pins arranged in a triangular lattice and forms an element of the experimental lead-based Versatile Loop Facility (VLF), developed by Ansaldo Nucleare and Westinghouse and currently in final phase of construction at the Ansaldo Nuclear's site in Wolverhampton, UK. In this study, the flow through several elements of the FPBS is computed, including a representative fully-developed bare subchannel arrangement, an inlet T-junction upstream of the main test section, and the full length of the main FPBS test section inclusive of spacer grids and instrumentation wires. Results indicate that although the upstream T-junction induces significant swirl that enters the main test section, the spacer grids function as an effective flow straightener. Further comparisons using different approximations of the turbulent Prandtl number revealed only minor differences in the predicted temperature distribution, as the low molecular Prandtl number of lead increases the influence of heat conduction relative to turbulent heat transport. Subsequently to the work presented in this paper, the CFD model developed will be benchmarked against experimental measurements collected in the VLF.

Westinghouse is developing its next generation of high-capacity nuclear power plants with the Westinghouse Lead-cooled Fast Reactor (LFR) (Ref. 1). The Westinghouse LFR is a pool-type 950 MWt (∼450 MWe Net) passively safe modular construction plant that contains all the primary system components within the reactor vessel, including the reactor coolant pumps (RCPs) and the primary heat exchangers (PHEs). Similar to the Westinghouse AP1000® Pressurized Water Reactor design, the Westinghouse LFR design incorporates passive cooling capability primarily through its Passive Heat Removal System (PHRS). During accidents, the PHRS allows decay heat to be transferred from the core to a pool of water surrounding the guard vessel, which transitions to long-term air cooling upon water depletion. All of this takes place without operator action or need for moving parts or power supply. Overall, when combined with natural circulation of liquid lead inside of the reactor vessel, this passive system fully relies on natural circulation, thermal radiation and boiling mechanisms to transport decay heat from the core to the atmosphere. In this study, complementary Computational Fluid Dynamics (CFD) and system code analyses were performed to assess the passive cooling ability of the Westinghouse LFR design during station blackout (SBO) conditions. The postulated scenario examined in this study is the longer-term phase of the SBO event, with the RCPs and PHEs assumed to be inactive and the water inventory in the pool exterior to the guard vessel fully depleted, thus relying purely on air cooling as the ultimate heat sink. The CFD analysis was conducted with Siemens STAR-CCM+ code using, as domain, the primary lead pool and with boundary conditions provided by the containment code GOTHIC. The CFD analyses provided high fidelity resolution of the three-dimensional natural circulation flow patterns formed throughout the reactor vessel. The results of the study not only provided confirmation of the passive cooling capability of the Westinghouse LFR design during an SBO event, but also provided granular input for vessel structural analysis activities accounting for thermal gradients developed during an SBO event.

The project “Modeling and Validation of Sodium Plugging for Heat Exchangers in Sodium-cooled Fast Reactor Systems” was conducted jointly by Westinghouse Electric Company (Westinghouse) and Argonne National Laboratory (ANL), over the period October 1, 2013- March 31, 2016. The project’s motivation was the need to provide designers of Sodium Fast Reactors (SFRs) with a validated, state-of-the-art computational tool for the prediction of sodium oxide (Na2O) deposition in small-diameter sodium heat exchanger (HX) channels, such as those in the diffusion bonded HXs proposed for SFRs coupled with a supercritical CO2 (sCO2) Brayton cycle power conversion system. In SFRs, Na2O deposition can potentially occur following accidental air ingress in the intermediate heat transport system (IHTS) sodium and simultaneous failure of the IHTS sodium cold trap. In this scenario, oxygen can travel through the IHTS loop and reach the coldest regions, represented by the cold end of the sodium channels of the HXs, where Na2O precipitation may initiate and continue. In addition to deteriorating HX heat transfer and pressure drop performance, Na2O deposition can lead to channel plugging especially when the size of the sodium channels is small, which is the case for diffusion bonded HXs whose sodium channel hydraulic diameter is generally below 5 mm. Sodium oxide melts at a high temperature well above the sodium melting temperature such that removal of a solid plug such as through dissolution by pure sodium could take a lengthy time. The Sodium Plugging Phenomena Loop (SPPL) was developed at ANL, prior to this project, for investigating Na2O deposition phenomena within sodium channels that are prototypical of the diffusion bonded HX channels envisioned for SFR-sCO2 systems. In this project, a Computational Fluid Dynamic (CFD) model capable of simulating the thermal-hydraulics of the SPPL test section and provided with Na2O deposition prediction capabilities, was developed. This state-of-the-art computational tool incorporates a first-principles Na2O deposition model developed by ANL, and combines it with predictive capabilities for the spatial and temporal variation of temperature, velocity, dissolved oxygen concentration, and wall temperature under flowing sodium conditions. The CFD model was validated under no-deposition conditions using experimental data collected with the SPPL, demonstrating the model’s capability to predict the thermal-hydraulics of the SPPL test section within the measurement uncertainty characterizing the SPPL instrumentation. The model’s deposition prediction capability was not, however, validated as the SPPL could not be operated under plugging conditions during the project, resulting in the lack of deposition data with adequate pedigree for a CFD model validation. Two novel diagnostic techniques to detect and characterize Na2O deposits, i.e. Ultrasonic Time Domain Reflectometry (UTDR) and Potential Drop (PD) techniques, were developed to ultimately assist in the validation effort under plugging conditions, which can be performed once the SPPL becomes operational. This development effort consisted first in demonstrating, analytically and/or computationally, the capability of these techniques to diagnose Na2O deposits inside of small channels (particularly the deposit’s thickness), and subsequently in the fabrication and testing of prototypical UTDR and PD instrumentation. The testing, performed on mockups of the SPPL test section, demonstrated the capability of these techniques to detect and characterize material discontinuities like those induced by sodium oxide deposition on stainless steel channel walls. Because of the mentioned impossibility to run the SPPL in a plugging mode, the developed instrumentation could not be tested in-situ, i.e. at the SPPL while deposits are being formed inside of the SPPL test section. Recommended future work includes a possible enhancement in the CFD modeling technique and installation of the developed UTDR and PD instrumentation on the test section, followed by plugging tests to be conducted with the SPPL. The installation of the UTDR and PD diagnostic instrumentation on the SPPL test section will allow collection of Na2O deposition data after the onset of deposition to nearly complete channel plugging, which can ultimately be used for the validation of the CFD model.