Liquid Metal Fast Reactor Research Articles

A subchannel analysis code for advanced liquid metal fast reactor cores and study on heat transfer characteristics of core geometry parameters

The liquid metal fast reactor represents an important reactor type for the future development of nuclear energy. Accurately modeling and predicting the subchannel flow and heat transfer phenomenon in the rod bundles plays a key role in ensuring core safety. Consequently, a subchannel code suitable for the liquid metal fast reactor is analyzed based on the subchannel analysis code of the Pressurized Water Reactor (PWR). The modified code incorporates various types of liquid metals, such as lead, lead-bismuth eutectic (LBE), and sodium, along with their constitutive models, enhancing the applicability of the liquid metal reactor systems. This code has been verified and validated at several levels, including comparing the code simulation results with other similar subchannel codes results, high-dimensional Computational Fluid Dynamics (CFD) simulations, and experiment data. The sensitivity analysis of core geometric parameters and turbulent mixing coefficients is performed based on the modified version code. The influence of the core geometry, including the rod Pitch to Diameter ratio (P/D), Rod to Wall gap (RTW) and the wire wrap pitch (H) on the sensitivity of outlet temperature has been investigated. The results show that the new subchannel analysis code suitable for liquid metal cooled reactor shows reasonable agreement with the verified and validated results. The geometric parameters, such as P/D, RTW and H have noticeable effects on the outlet temperature difference of the subchannels. Additionally, the outlet temperature distribution in the subchannel is significantly affected by different turbulent mixing coefficients and the distribution of the turbulent mixing coefficient also influenced by the reactor core sizes. Overall, this study could support the rod bundles design and safety analysis in the liquid metal reactors.

Development of a liquid metal subchannel code applied to ocean conditions and study on the effect of core geometry parameters and ocean motions on flow and heat transfer characteristics

The Generation IV International Forum (GIF) proposes a type of liquid metal reactor, due to its thermal properties and efficient heat transfer in a relatively small system. This special heat transfer characteristic allows the liquid metal reactor to be modular in design and flexible in installation on a nuclear-powered ship. Accurate prediction of the coolant temperature and flow distribution between subchannels is important to ensure nuclear safety in the complex ocean environment. In this study, an ocean version of subchannel code is developed by adding the motion terms in the momentum equations for the liquid metal reactor core based on the previously land-based subchannel code. Since the research on the characteristics of flow resistance and heat transfer mechanism under ocean conditions is not mature, the original empirical model is still used in the ocean version code. The constitutive models of the liquid metal are analyzed and incorporated into the ocean code. The code has been verified and validated by comparing with Computational Fluid Dynamics (CFD) simulation results, and experimental data. The effect of geometric parameters such as rod Pitch to Diameter ratio (P/D), Rod to Wall gap (RTW) is studied in the liquid metal lead 7 rod bundles. In addition, the effect of heaving start-up and shutdown motions between liquid metal lead and water is discussed in the 5 × 5 bare rod bundles. In general, the modified subchannel analysis code can well complete the calculation of liquid metal reactor under ocean conditions, as well as provide support for the design and safety analysis of a liquid metal cooled fast reactor.

Steady and transient analysis on thermal hydraulic characteristic of helical coiled once-through steam generator of liquid metal fast reactor

A transient model for thermal–hydraulic characteristics of helical coiled once-through steam generator (HCOTSG) of liquid metal fast reactor was established based on global discrete grid method. The model meshed the liquid metal circuit and water-steam circuit. Meanwhile, unsteady heat conduction equations were built to accurately simulate coupled heat transfer between two sides. Four-equation drift-flux method was adopt for water-steam flow. Taking lead–bismuth fast reactor as example, thermal–hydraulic characteristics of HCOTSG were first calculated under steady conditions. It was found that heat flux distribution along steam generator is extremely uneven, and the difference between maximum and minimum wall heat flux reaches hundreds of times. Then, the transient response was simulated when inlet conditions of primary side were perturbed by step changes, and it was found that maximum wall heat flux increases from 1175.5 kW/m2 to 1365 kW/m2, increasing by nearly 16 %, when inlet lead–bismuth temperature step increases from 450°C to 480 °C.

Investigation of the fast flux test facility transient behavior during a loss of flow without scram test

A Coordinated Research Project (CRP) on the benchmark analysis of Fast Flux Test Facility (FFTF) Loss of Flow Without Scram (LOFWOS) has been held under the sponsorship of the International Atomic Energy Agency. The CRP aims to improve understanding of loss of flow events in fast reactors, as well as to assess capabilities of existing computer codes against experimental data. FFTF is a 400 MWth Sodium-cooled Fast Reactor (SFR) owned by the U.S. Department of Energy, operated from 1980 to 1992. The CRP involves the LOFWOS Test #13, belonging to a series of unprotected transients performed as part of the Passive Safety Testing program. In this framework, the Nuclear Engineering Research Group of Sapienza University of Rome has developed an integrated multiphysics modelling based on a coupled approach with RELAP5-3D© (thermal-hydraulics) and PHISICS (neutron-kinetics) codes. The present paper discusses the results obtained by the authors after the participation in the benchmark open phase. As suggested by the organizers, a two-step methodology was followed. Firstly, calculations were run by imposing the reactor power as boundary condition and focusing on the FFTF thermal–hydraulic behaviour. A good accordance was detected between the numerical results and the experimental data, above all for what concerns the reference core outlet temperatures. Then, the simulations were repeated by using a coupled approach. Calculation outcomes demonstrated the capability of the selected code suite (PHISICS/RELAP5-3D®) to reproduce the reactor transient behavior and its suitability to be used as an effective numerical tool to simulate liquid metal fast reactors accidental scenarios where thermal–hydraulic and neutron-kinetic phenomena are strongly coupled.

Fluid Dynamics Effects of a Porous Blockage on Laminar Flow in the Interior Subchannels of a 61-Pin Wire-Wrapped Rod Bundle

Flow blockages in Liquid Metal Fast Reactor (LMFR) fuel assemblies that can cause fuel pin cladding failures require further investigation for the development and optimization of these Generation IV reactor designs. The objective of this study was to experimentally evaluate the effects of a porous blockage accident scenario on laminar flow behavior in the interior subchannels of a prototypical 61-pin wire-wrapped rod bundle. The laminar flow condition is considered because of its importance to natural circulation and loss-of-coolant-accident operating scenarios as well as the limited availability of experimental data. The matched-index-of-refraction method was utilized to obtain two-dimension two-component time-resolved particle image velocimetry (TR-PIV) measurements at three planes centered on blocked and neighboring subchannel regions. Time-averaged first-order and second-order statistics, computed by Reynolds decomposition, were visualized including the mean and fluctuating streamwise and spanwise velocity components. Line profiles described the evolution of flow from upstream regions, to separated flow, and recombination downstream, while a modified Galilean decomposition distinguished differences between flow in the blocked measurement plane and flow in its neighboring counterpart region. Spatial-temporal cross-correlations were performed for the streamwise velocity fluctuations to characterize the convection velocity and decay of traveling vortices in the wake of the porous blockage. The isothermal TR-PIV measurements from this study provide high-fidelity experimental data sets for the validation of computational models and numerical studies to characterize complex fluid phenomena in wire-wrapped rod bundles during the potential accident scenario of a porous, interior flow blockage.

Experimental study on practical application of optical fiber sensor (OFS) for high-temperature system

This study explores the application of Raman scattering-based optical fiber sensors (OFSs) in extreme environments, specifically focusing on a loop heater vessel with temperatures ranging from 200 °C to 680 °C. This condition generally covers the advanced reactor designs, such as Sodium-cooled Fast Reactor and High Temperature Reactor. Various optical fiber combinations were employed for temperature measurements, taking into consideration the operating temperature of the target equipment. Two types of OFSs, gold-coated and polyimide-coated, were utilized. Protective tubes made of stainless steel (STS) and carbon were introduced to ensure reliable temperature data collection in high-temperature settings. Results indicate that the STS tube with a gold-coated OFS exhibited the highest consistency and agreement with thermocouple measurements, making it suitable for extreme environments. The study emphasizes the applicability of this system in high-temperature environments, such as liquid metal reactors, high-temperature thermal energy storage system, and hydrogen production system, for environmental monitoring.

SIMMER code two-fluid model well-posedness and stability for simulating liquid metal (lead–lithium eutectic), water vapor, and non-condensable gases multi-component multi-phase flows

Liquid metals are employed as a coolant in liquid metal fast reactors (LMFR) and are considered breeders and coolants for future power-producing fusion reactors. The interaction of liquid metals with other coolants, such as air and water, is one of the possible occurrences in liquid metal-cooled reactors. Although the SIMMER-III computer code is currently in the validation and verification phase, it is a prospective candidate code for correctly investigating possible accidents. This study aims to determine the stability and well-posedness of the SIMMER-III code Eulerian-Eulerian two-fluid model (TFM) under any such accident scenario. This work also considers the influence of the virtual mass force and diffusion forces in momentum on TFM stability in accelerated liquid metal, steam, and non-condensable gaseous flows throughout all flow regimes. The characteristics method is used to assess the ill-posed nature of TFM for all types of accidents and accident scenarios in a hypothetical simplified system model with several components (Lead–Lithium, non-condensable gases, and water vapor). It has been discovered that the analysis findings vary from the air and water two-component two-phase flows (characteristics roots spectrum and error growth rate patterns) and are particularly sensitive to the diffusion and virtual mass coefficients. This is because liquid metals have a higher density than liquid water and steam, resulting in strong virtual mass forces and weak diffusion forces in liquid-metal and gas two-phase flows. Because of the extensive range of possible interactions between fluxes of different components, producing an accurate representation of diffusion in multi-component mixtures is difficult. Because of this, it is strongly suggested that the virtual mass coefficient and diffusion coefficients be handled more accurately for these kinds of flows. The values of these coefficients significantly affect how accurate proposed TFM predictions are, but there hasn't been much research on how to estimate them. The study also sheds light on the model's accuracy and highlights the areas where the model's predictions will be mathematically trustworthy.

LES of the upper plenum of the liquid metal fast reactor under the forced flow conditions

A Large Eddy Simulation (LES) of the upper plenum of the E-SCAPE (European SCAled Pool Experiment) facility under the normal operations has been conducted to gain deeper insights into the thermal hydraulics phenomena in liquid metal fast reactors (LMFRs). The results unravel the overall flow features in the upper plenum, the thermal instability in the above core structure region and the impact of the jets from the barrel holes on the large flow circulation and thermal mixing.The above core structure region is represented using a homogeneous porous model. It is shown that during the normal operations, most of the hot stream of the lead bismuth eutectic (LBE) from the core center rises to the top and disperses from the upper barrel holes. Conversely, the cold LBE spread out from the lower barrel holes. Mixing between the two streams only affect the fluids leaving from the middle heights demonstrating limited mixing. These jets help to create a large circulation in the upper plenum region which prevents any thermal stratification. At the interface between the hotter and colder streams, there are unsteady large-scale structures resulting in a mixing layer in the thermal field. This strong mixing is not conventional turbulence and potentially cannot be captured by Reynolds Averaged Navier Stokes average (RANS) modelling.The flow in the upper plenum is dominated by the influences of the different types of jets, which include some free jets issued horizontally or angled upwards and some jets impinging onto the structures in the plenum. The jets overall behave similar to ‘standard’ jets, though significant interactions between the top and bottom jets and the background circulations have strong influences at later stages of the jets, typically after four jet diameters leading to the jets not reaching self-similarity. In addition, the turbulence that is observed in the upper plenum is largely generated by the jet flows and hence the distribution of turbulence is very non-uniform. Away from the jets, turbulence is minimal with the mixing largely driven by the large-scale circulation.

METAL: Methodology for liquid metal fast reactor core economic design and fuel loading pattern optimization

The design of Liquid Metal-cooled Fast Reactor (LMFR) cores encompass two levels of design. The first is the physical core design, which is concerned with the design of the fuel assembly geometry in the context of a full core. The second is the fuel loading design, which is an in-core fuel management problem concerned with the design of fuel enrichment and core loading pattern. In this paper, an optimized fuel design search is developed to improve core loading patterns. As part of the implementation, the multiphysics simulation suite LUPINE has been enhanced to allow for fuel shuffles and an in-core fuel management optimization methodology has been added with the objective of reducing the fuel cost. The design methodology is referred to as Methodology for Economical opTimization of Applied LMFR (METAL), and this methodology is demonstrated using the popular Super Power Reactor Innovative Small Modular (SPRISM) sodium-cooled fast reactor (SFR) as a reference core. One challenge currently facing the deployment of LMFRs is the availability of TRansUranium (TRU) and high assay low enriched uranium (HALEU) driver fuel. The two forms of fuel are expensive and not widely available, which poses a challenge on the startup of LMFR cores. To address the availability of driver fuel, and to lower fuel costs, low enriched uranium (LEU) blankets are investigated as replacements to the natural uranium or depleted uranium blankets typically used. The advantage of this solution is the wide availability of LEU and its ability to lower the amount of TRU or HALEU driver fuel needed. The design objectives, constraints, and optimization algorithm for a SFR are identified. Then, METAL is applied to design a SFR core fueled by TRU driver fuel assemblies and LEU blanket assemblies. To demonstrate the advantage of introducing LEU blankets, METAL is also used to design another SFR core following the same objectives and constraints, but using naturally enriched blankets. By comparing the two cores, it was found that the introduction of LEU blankets results in a ∼28% reduction in the driver fuel mass requirements. Upon development of a levelized fuel cycle cost (LFCC) model, the 28% reduction in driver fuel mass corresponds to a 10% decrease in the LFCC. Additional advantages of using LEU blankets include reducing the core conversion ratio (CR), and improving the assembly radial power peaking factor (RPPF), which enhances the safety and non-proliferation performances of the SFR core.

Study on the turbulent Prandtl number model for liquid metal flow and heat transfer in a narrow rectangular channel

In order to improve the core power density of liquid metal reactors, plate-type fuel assemblies are typically used, which have obvious advantages in heat transfer capacity. For the low Prandtl number fluid, the RANS turbulence model-based CFD method needs to be modified by inserting the turbulent Prandtl number model. This paper investigated the turbulent Prandtl number model for the turbulent heat transfer of liquid metal flow in a narrow rectangular channel. First, the existing turbulent Prandtl number models were summarized and classified as the local turbulence Prandtl number model and the global turbulence Prandtl number model. They were evaluated based on the experimental data. The results showed that the prediction results of each model are not ideal. Then, a typical numerical example was used to analyze the temperature field and flow field in the narrow gap of a narrow rectangular channel. It was shown that the entire flow region only includes the near-wall region and the transition region, and the transition region accounts for the main proportion. A semi-empirical local turbulence Prandtl number model was obtained by solving the turbulent transport equation, and the unknown coefficients were determined iteratively with the experimental data. The error was controlled within ±20%. Finally, the applicability of the newly proposed turbulent Prandtl number model was explored, and the results showed that it has good applicability for the different runner sizes and different liquid metal types. This study can provide a reference for the numerical simulation of turbulent heat transfer of metal liquid in the narrow rectangular channel.

Experimental study on fluid-structure interaction of multiple support barrels in liquid metal fast reactor

In an earthquake, the main equipment support barrels in the pool-type liquid metal fast reactor (LMFR) may experience strong vibrations due to the fluid-structure interaction (FSI) phenomenon. Accurate parameters for FSI characteristics of multiple cylinders are crucial for the seismic design and analysis of these support barrels. Limited experimental data is available for pressure characteristics and FSI parameters, particularly for closely distributed thin-wall internal support barrels. In this research paper, we conducted a mechanism experiment for FSI parameters on support barrels in the fast reactor by performing shaking table tests. We analyzed the variation of the added mass coefficient with different height-diameter ratios and distances (gaps) and studied the pressure distribution and added mass characteristics for the seismic test. According to the study, the cylinder's added mass coefficient is higher under seismic conditions compared to the experimental results obtained from frequency sweeps. This suggests that in addition to the modal additional mass obtained by the modes, the impact of inertial effects may also need to be considered under seismic conditions. We then added the obtained added mass to the finite element model to conduct transient seismic simulations. The results show that the maximum error between experiments and simulations is 15%, which is considered reasonable for approximate engineering applications.

Review of fission gas release in liquid metal reactor fuel cladding failure accident

The release of gas fission products from liquid metal fast reactors in the event of cladding failure accidents is a significant focus during the development and improved design of such reactors. Researchers worldwide have devoted considerable attention to this topic. But studies aiming at liquid metal reactors have not yet reached the same level as studies of fission gas release in water-cooled reactors. To address this gap, this paper reviews the research progress of related fission gas release from the aspects of mechanism model, experimental research, numerical simulation, and measurement. Considering the potential essential similarity of the release process of fission gas in water-cooled reactors and liquid metal cooled reactors, the researches on water-cooled reactors are also included. This review helps to understand the behavior and characteristics of fission gas release in liquid metal cooled reactors and provides some guidance for subsequent numerical simulation and experimental research on fission gas release based on liquid metal cooled reactors.

Study on flow and heat transfer characteristics of liquid metal flow in narrow rectangular channels under high heat flux

In order to improve the core power density of liquid metal reactors, narrow rectangular fuel assemblies are typically used, which have obvious advantages in heat transfer capacity. However, the flow and heat transfer characteristics in coolant passage need to be studied. In this article, the numerical simulation analysis model obtained earlier was applied to calculate the Nusselt number, friction factor, with different Reynolds number, Prandtl number, or Peclet number of lead bismuth flow in a narrow rectangular channel under high heat flux. A heating correction factor was proposed and a flow resistance model was established for high heat flux. The influencing factors of ultra-high flux reactor core, such as gap size, flow direction, single and double sided heating conditions, on the flow and heat transfer in narrow rectangular channels were explored. The results showed that gap size and aspect ratio have a certain impact on the Nu and f of narrow rectangular channels, but the influence is relatively small within the narrow rectangular channel condition range. The influence of flow direction on flow and heat transfer is minimal, the single and double sided heating conditions will bring about a difference of about 20% in Nu and a difference of about 2.5% in f. In addition, the type of liquid metal will also affect the results. The research can provide reference for the thermal design and optimization of ultra-high flux reactor cores.

Study on coupled heat transfer model and enhanced heat transfer law of liquid metals and supercritical fluids

ABSTRACT The supercritical carbon dioxide Brayton cycle system is being considered for application in the power conversion system of advanced liquid metal fast reactors. One important research topic is the coupled heat transfer characteristics and enhanced heat transfer laws between liquid metals and supercritical fluids. However, the traditional Reynolds analogy hypothesis with a constant turbulent Prandtl number is difficult to satisfy the numerical conjugate heat transfer research of liquid metals and supercritical fluids. Therefore, based on the open-source computational fluid dynamics program OpenFOAM, a numerical method suitable for the coupled heat transfer calculation of liquid metals and supercritical fluids was developed, which can achieve precise conjugate heat transfer solutions by applying different turbulence models and heat flux models to liquid metals and supercritical fluids. Subsequently, based on this method, the coupled heat transfer behavior of liquid metals and supercritical fluids in the straight channel of a typically printed circuit heat exchanger was studied. The focus was investigating the conjugate heat transfer characteristics of liquid heavy metal lead-bismuth eutectic and liquid metal sodium with supercritical carbon dioxide. Some sensitive results, including the inlet temperature and Reynolds number of liquid metals and supercritical fluids, were qualitatively and quantitatively analyzed.

Design Optimization and Measurement Uncertainty of an Electromagnetic Level Sensor for Liquid Metal Reactors

Design Optimization and Measurement Uncertainty of an Electromagnetic Level Sensor for Liquid Metal Reactors

In-situ kinetic study of irradiation induced crystallization in amorphous Al2O3

In the last ten years amorphous alumina coatings, deposited by Pulsed Laser Deposition, emerged as potential key enabling technology in the fields of heavy liquid metal fast reactors (lead and lead-bismuth) and fusion. In the former, as coating of the steel fuel cladding and in the latter as multifunctional coating providing a barrier against tritium permeation, steel corrosion and electrical insulation. Nevertheless, a detailed knowledge of the behavior of this thermodynamically metastable material at high temperatures and under neutron irradiation is still unknown. A knowledge gap that is mandatory to fill up for the deployment of this barrier technology. In the present work, we present a first step towards this goal, by the in-situ dynamic observation of the radiation induced crystallization processes of thin films of amorphous Al2O3, induced by ion-irradiation over an extensive range of temperatures (400-800 °C). The study was performed at the Intermediate Voltage Electron Microscope (IVEM)-Tandem Facility at Argonne National Laboratory. The experimental findings allow to elucidate the dependence of the grain growth on ion dose and temperature. A kinetic approach has been used to derive the process activation energies and other important parameters.

Experimental characterization of pressure and friction factor in an interior subchannel of a 61-pin wire-wrapped rod bundle with a porous blockage

Comprehending and counteracting accident conditions presented by impedances of flow in diminutive subchannels of a Liquid Metal Fast Reactor (LMFR) hexagonal rod bundle are imperative toward their development and safety. Scarce experimental research currently exists in the literature to characterize the pressure and friction factor for partial blockages in LMFR assemblies. Experimental pressure measurements were conducted in a 61-pin prototypical LMFR fuel assembly using specialized instrumented wire-wrapped rods with a three-dimensional printed porous blockage installed. The pressure drop was measured for one helical pitch at four distinct interior subchannel locations: two in the blocked subchannel and two unblocked adjacent locations (near-center and near-wall of the assembly). A wide range of Reynolds numbers between 140 and 24 000 were studied to evaluate the blocked subchannel friction factor and to determine the flow regime boundaries for laminar-to-transition and transition-to-turbulent flows. Power spectral density analysis of the pressure fluctuations for three distinct locations (one upstream and two downstream of the porous blockage) revealed the mechanisms of coherent structure formations and transport, and dominant location-dependent Strouhal numbers. One-dimensional continuous wavelet transforms of the pressure fluctuations demarcated temporal instances of flow events with their frequency content. Temporal cross correlation quantified the temporal delay between the blocked subchannel pressure fluctuations in the blockage vicinity. The presented research provides first-of-its-kind datasets and fluid physics based-analyses for the interior LMFR subchannel in the presence of a porous blockage and provides a benchmark for the validation of computational flow models and predictive correlations for the safety enhancement of LMFR rod bundles.

Development of a clogging probability model in simplified geometry for fast reactor flow blockage analysis

The occurrence of flow blockage accident in fuel assemblies is considered as one of the primary challenges to be addressed in liquid metal fast reactors. Flow blockage accidents are primarily caused by the accumulation of solid particles, such as corrosion products, wire spacer fragments, and chemical products between liquid metal and impurities like air or water. To better understand the mechanisms underlying the blockage formation, a project consisting of several series of simulated experiments is currently underway at Sun Yat-sen University and Harbin Engineering University. In the initial stage of the project, experimental study was conducted to investigate the influence of particle characteristics on the probability of clogging by discharging various solid particles into a reducer pipe. Preliminary experimental analyses revealed the formation of an arch structure at a specific location on the pipe. This arch structure causes the granular flow to cease, resulting in blockage. Based on the accumulated experimental database, in this study an empirical model is developed to predict the clogging probability, taking into account the above-mentioned influential parameters. The model demonstrates a reasonable agreement with experimental results within the current range of conditions, encompassing variations in the ratio between hopper outlet size and particle size, particle capacity, particle static friction coefficient, and particle release height. Although further extension of the empirical model is inevitably necessary under more realistic reactor conditions, such as incorporating realistic sub-channels, irregularly shaped particles, or a loop with a liquid phase, it is expected that the model will be beneficial for the improved design of sub-channels and safety analysis of fast reactors.

State-of-the-art review of the T/H system codes RELAP5 for HLM applications

Thermal-hydraulic code RELAP5, originally developed by INEEL under NRC contract for LWRs LOCA analysis, has been extensively validated and it is actually used worldwide as a best estimate code for LWRs transient analysis. This code was chosen in the early 2000s as the reference code for the thermal-hydraulics and accident analysis of Heavy Liquid Metal Fast Reactors within the frame of Italian research programs. Several original RELAP5 routines have been updated in order to implement suitable correlations generating the reference physical and thermodynamic properties for Lead and Lead-Bismuth Eutectic fluids. Similarly, some specific heat transfer correlations for liquid metals have been added. In almost twenty years, ENEA has conducted an extensive validation program to demonstrate the capability of the code to simulate the thermal–hydraulic behaviour of heavy liquid metal systems. This program was mainly conducted on the facilities of the Brasimone center (CHEOPE, CIRCE, NACIE), but also used data from international programs (MEGAPIE International Experiment and LACANES Benchmark OECD/NEA). Furthermore, the participation of ENEA in the projects financed by the European Commission in several Euratom programs for the study and design of HLM-cooled reactors (ELSY, SEARCH, MAXSIMA, LEADER, SESAME. MYRTE) has allowed to increase the confidence on the transient and accidental analyses conducted with RELAP5 also through the comparison with the results of other codes. RELAP5 is currently a reliable tool both as a support to reactor design (safety by design) and for accidental analyses.

Assessment of RANS and LES in predicting the flow in a 61-pin wire-wrapped bundle with a porous blockage

One of the most promising nuclear reactor designs is the Liquid-Metal Fast Reactor (LMFR). Its fuel bundle rods are separated with wire spacers that are helically wrapped around each fuel pin. Since this bundle presents a tight rod pitch to diameter ratio, accumulation of debris, cladding deformation or detachment of the wires may generate the formation of porous blockages. Understanding the coolant behavior in LMFR fuel bundles under the influence of blockages is mandatory for a complete nuclear safety evaluation. Therefore, this paper presents Computational Fluid Dynamics (CFD) results using the RANS k-ε Realizable and the LES models to simulate the flow of a 61-pin wire-wrapped bundle with the presence of a porous blockage that obstructed six interior subchannels. The simulated Reynolds numbers were 5,000 (transition regime) and 14,000 (turbulent regime). Results of the first and second order velocity statistics were compared with the Particle Image Velocimetry (PIV) experimental data acquired in the Thermal-Hydraulics Research Laboratory facility of Texas A&M University. The simulations results were found in reasonable agreement with the PIV data. Additional analysis of the flow structures, pressure distribution and power spectral densities were also performed to give more insight of the flow characteristics in LMFR reactors under the presence of porous blockages.