Liquid Metal Fast Reactor Research Articles

Advanced Thermal-Hydraulic experiments and instrumentation for heavy liquid metal reactors

Heavy-liquid metals (HLMs), such as lead and lead–bismuth eutectic (LBE), are proposed as primary coolants in accelerator driven systems and next-generation fast reactors. In Europe, the reference systems using HLMs are MYRRHA (LBE) and ALFRED (lead). Extensive R&D programs have been established for supporting their detailed design and safety assessment, including thermal–hydraulic experiments at representative operating conditions in an HLM environment. These experiments aim both at a design verification and at the validation of numerical models, which allow an extrapolation of the results. Advanced instrumentation, capable of sustaining high temperatures and corrosion, is necessary for accurate measurements, often in compact geometries.This article presents an overview of recent experiences and ongoing activities on pool-type and loop-type HLM experiments. Pool tests include the measurement of forced- and natural-circulation flow patterns in several scenarios representative of nominal and decay heat removal conditions. Loop tests are focused on the evaluation of specific components, like mockups of the fuel assembly, control rod and heat exchangers. They involve the measurement of global variables, such as flow rate and pressure difference, and local quantities like temperature, velocity and vibrations. In addition to traditional techniques, other instrumentation based on optical fibers, ultrasonic and electromagnetic methods are discussed.

Computational Analysis of Thermal Striping in Primary Sodium System of Liquid Metal Fast Breeder Reactor Using Finite Volume Method

Thermal striping is associated with random fluctuations of temperature that occur at the nonisothermal jet stream interface or across thermally stratified fluid layers due to the high heat transfer coefficient of liquid sodium flow. The temperature fluctuations in the jet mixing or stratified layer regions are transmitted to the adjoining structures after minimal attenuation in a Liquid Metal Fast Breeder Reactor (LMFBR). In turn, the adjoining structure may experience high cycle fatigue and catastrophic failure caused by crack propagation. Investigations have been carried out in detail numerically, and frequency and amplitude of temperature fluctuations in 500-MW(electric) pool-type fast reactor [Prototype Fast Breeder Reactor (PFBR)] structures for practical applications have been observed. The investigations consist of numerical simulations at two levels. First, a published benchmark experiment is analyzed, and then, a suitable computational fluid dynamics (CFD) model is identified for simulating the thermal striping phenomenon numerically. After that, detailed flow and temperature fluctuations are predicted in the reactor structures by analysis carried out based on the CFD model. The values of the temperature fluctuations predicted are found to be within acceptable limits, as required by structural mechanics considerations in the study.

Analysis of the turbulent flow in a partially blocked wire-wrapped rod bundle using LES with wall functions

One of the most promising nuclear reactor designs is the Liquid-Metal Fast Reactor (LMFR). Its fuel assembly consists of rods containing the fuel material, arranged in a triangular lattice and covered by a hexagonal duct. The rods are separated with wire spacers that are helically wrapped around each fuel pin. Because this geometry is tight and compact, accumulation of debris, cladding deformation or detachment of the wires may generate a partial or total flow blockage of coolant subchannels at different locations. The understanding of the coolant behavior in LMFR fuels under the influence of a partially blocked flow area is required for a thorough safety assessment. A Large Eddy Simulation (LES) with wall functions was executed to simulate the flow on a 61-pin wire-wrapped rod bundle under the presence of a blockage located on the bundle periphery to simulate an abnormal operation condition. The Reynolds number was Re=17000, which is in the turbulent regime. Results of mean velocity and root mean square (rms) fluctuating velocity profiles were compared with PIV experimental data from the Texas A&M Thermal-Hydraulic Research Lab. Both mean and rms velocities were found in reasonable agreement with the experimental data among different locations because the simulation results were within 2σ from the PIV measurements. A pressure form loss coefficient due to the blockage was calculated and found to be Kloss=1.0455. The fluctuating velocities revealed a pair of shear layers that formed on each side of the blockage carrying the highest turbulent intensities. A spectral analysis was conducted to visualize the dominant frequencies on the turbulent flow in the recirculation region. The spatial–temporal two-point cross-correlations based on the fluctuating velocities were calculated at different points along the shear layer and were used to estimate integral lengths scales and convection velocities. The cross-correlations showed larger structures at higher axial elevations from the blockage rear wall. The length scales were found in the range of 3 to 21 mm while the magnitudes of the convection velocities presented values between 0.69U¯in to 1.39U¯in, where U¯in is the inlet velocity. Proper orthogonal decomposition (POD) analysis was utilized to extract the dominant modes in the fluctuating flow. Large flow structures and low frequency time signals were observed for the lowest order modes.

Assessment of nuclear data libraries for fast reactor systems with MCNP and NDaST

The paper presents results for multiplication factors of reactor physics experiments with fast spectrum systems, representative of liquid metal fast reactors, documented in the IRPhEP Handbook. The calculations were performed with the Monte Carlo code MCNP 6.1, using continuous-energy data based on ENDF/B-VII.1, ENDF/B-VIII.0, JEFF-3.1.2, JEFF-3.2, and JEFF-3.3. Considerable discrepancies were observed between the results obtained with the different data libraries. In comparison with the benchmark values of the experiments, the ENDF/B-VII.1 results show the best agreement, while for the JEFF-3.3 results, the deviations are largest. In addition, the sensitivity tool NDaST, using linear perturbation theory, was applied to identify the most important microscopic cross sections responsible for the differences of the latest library revisions ENDF/B-VIII.0 and JEFF-3.3 as compared to ENDF/B-VII.1. For JEFF-3.3 vs ENDF/B-VII.1, where the deviations are overall largest, the main contributors are U-238n, γ and, for systems containing plutonium, Pu-239 fission and total neutron multiplicity.

Development and preliminary validation of fully implicit scheme single-phase liquid metal fast reactor code NUSOL-LMR

A new simulation code (NUSOL-LMR) is developed to simulate the thermal–hydraulic response to hypothetical accidents in liquid metal fast reactors. The basic field equations of NUSOL-LMR consist of the single-phase continuity equation, momentum equations and energy equations. Heat conduction model of heat structure is also included. A Newton-based fully implicit numerical scheme is used to overcome the courant time step limitation. Reasonable accuracy and no stability limitation with fast-running are confirmed by comparing results from NUSOL-LMR with experimental data and results from other codes.

Demonstration of the model of physicochemical processes in heavy liquid metal cooled reactors in an example of the THEADES experiment simulation

Activities have been currently underway across the world to build heavy liquid metal-cooled reactors. Thus, for example, the ALFRED and BREST-OD-300 reactor designs use lead coolant, while the MYRRHA and CLEAR-I reactors use lead–bismuth coolant. The high corrosive and erosive activity of the coolant requires regulations oxygen concentration (1 ÷ 4E-6 wt.%) and flow velocity (0.5 ÷ 2.5 m/s). Due to the complex geometry of the reactor circulation circuit flow paths, numerical simulation methods are used extensively to justify scheduled modes of operation for liquid metal coolants. Correct justification of the complex oxygen transport processes in liquid metals requires the respective physicochemical computation model, which takes into account the main reactions of oxygen with the coolant and the structural materials.This paper presents a physicochemical model, which includes the following processes: erosion, growth, and dissolution of the two-layer oxide film, coagulation and dissolution of metal oxides in the circuit with subsequent deposition on filtering elements, and inflow from mass transfer apparatuses. StarCCM+, a commercial CFD code, was used as the tool in this study. The physicochemical model was implemented using models of passive impurities, which are used to simulate oxygen transport in the circulation circuit.Capabilities of the presented model were demonstrated based on the results of investigating thermohydraulic and physicochemical processes obtained at the THEADES experimental facility. The duration of the simulation was 1000 h.The distributions of impurity concentrations, increased erosive activity areas, as well as the total amount of oxides deposited on filter and the amount of oxygen entering the circuit have been obtained as a result of the calculations. The surface distribution of the oxide film thickness on the test facility surfaces contacting the coolant has also been calculated. These data can be extrapolated to reactor elements that operate under conditions that are similar to the considered experiment. For example, fuel assemblies.Simulation of thermocouples, as well as taking into account manufacturing technology made it possible to improve the accuracy of calculating the flow’s thermohydraulic characteristics as compared to earlier studies.

Thermophysical Characteristics of Liquid Metal In-Bi-Sn Eutectic (Field's Metal) as a Similarity Coolant

Abstract This work investigated the thermophysical characteristics of liquid indium–bismuth–tin eutectic alloy also known as Field's Metal for the purposes of use as a similar fluid for liquid metal reactors. The density, specific heat capacity, viscosity, thermal conductivity, and the coefficient of thermal expansion were determined for liquid Field's Metal for temperature ranges from its melting point 333 K to 423 K. The work captured the effect of temperature on these properties and each property's magnitude. The findings were used to create mathematical correlations to predict the value of the thermophysical property at a specified temperature for use in natural circulation studies. Notably, the work also observed non-Newtonian shear thinning behavior of the alloy near the melting point and that the non-Newtonian behavior relaxes as the material obtains more energy. The results are consistent with the behavior of other liquid metals including variances that occur close to the melting point.

Study on the flow and heat transfer characteristics of liquid sodium in a hexagonal 7-rod bundle

The flow and heat transfer characteristics of liquid metal sodium are different from the ordinary fluids, which brings great challenges to the CFD simulation in the sodium-cooled fast reactor (SFR). Therefore, the accurate flow and heat transfer models are particularly significant in the high fidelity three dimensional analysis of liquid metal reactors. In this paper, the OpenFOAM solver implemented with k – ε – kθ – εθ turbulent heat transfer model was developed and used to study the flow and heat transfer characteristics of liquid metal sodium. Firstly, the proposed thermal hydraulic models were validated against the thermal hydraulic experiment of single-phase liquid sodium in a 7-rod bundle. The key thermal hydraulic parameters, such as the Nusselt number, pressure drop and frictional coefficient, were compared with the experimental data. Results show that the developed OpenFOAM solver with new implemented model performs well and the simulated results were in good agreement with the experimental data. Then the thermal–hydraulic characteristics of liquid sodium in the bend rod bundle were investigated. The effects of bending angle, Reynolds number on peak wall temperature and Nusselt number were investigated. CFD results suggest that the heat transfer in the bend rod bundle is highly sensitive to the bending angle but not to Reynolds number. Two opposite-direction cross flows were formed in the sections before and after the maximum bend position, which reveals the formation and disappearance of hot spot in the narrow gap.

On steady state noncondensable gas buildup in Autonomous Reactivity Control (ARC) systems

The Autonomous Reactivity Control (ARC) system, a new engineered passive safety system for liquid metal reactors, has been the subject of many recent studies, having shown the potential to improve both safety and performance of liquid metal reactors while introducing only a minor impact on normal operations. This study investigates one area of ARC system performance which has not yet been examined — the production of internal noncondensable gases during normal operations. Using neutronic calculations, this study finds that a non-negligible quantity of gases will be produced through transmutation reactions of the internal ARC system fluids, and a lower bound on this quantity is calculated. Through a series of geometric arguments and considerations on mutual solubility and chemical reactivity, the volume and distribution of noncondensable gas buildup is examined for a wide range of ARC system designs. To demonstrate the impact of these noncondensable gases, a series of thermal-hydraulic and transient calculations are performed, and it is shown that the production of noncondensable gases from transmutation of the liquid potassium expander fluid will have negative impacts on the performance of the ARC system, causing an unsteady and restricted actuation which will lead to a reduction of benefits for a given reactor’s transient performance and an increased uncertainty on the state of the system during reactor operations. In the end, multiple mitigation strategies are discussed, but ultimately this study motivates a reevaluation of the fundamental ARC system design so that its benefits may be most fully realized.

Development and validation of a new oxide fuel rod performance analysis code for the liquid metal fast reactor

Development and validation of a new oxide fuel rod performance analysis code for the liquid metal fast reactor

Code-to-code SIMMER/FRENETIC comparison for the neutronic simulation of lead-cooled fast reactors

This paper presents a neutronic benchmark between the FRENETIC and the SIMMER codes, performed considering the core of the ALFRED lead-cooled fast reactor. FRENETIC includes a nodal diffusion module developed for fast full-core analyses of safety-relevant transients in liquid-metal fast reactors, while SIMMER is a reference tool for qualifying such kind of systems in accidental conditions. In order to highlight the influence of the different numerical methods available in the two codes on the output responses, macroscopic parameters like the effective multiplication factor and the linear power are compared in an increasingly detailed simulation framework. The 3D full-core FRENETIC results are compared at first to a cylindrical 2D SIMMER core model, to highlight the impact of the absorber rings location. Then, a 3D full-core model is also employed in SIMMER to assess the differences, with respect to FRENETIC, induced by the adoption of a Cartesian mesh and of the discrete ordinates method. The full input data consistency is ensured providing the same set of few-group material constants, generated with the Monte Carlo code Serpent 2. Finally, the W-ε scheme is applied to accelerate the angular and spatial convergence of the keff sequences computed by SIMMER in order to estimate the asymptotic errors with respect to the reference Serpent calculation.

Void effect research on MOX-fueled lead-cooled fast reactor

Liquid metal-cooled reactor (LMR), including lead-cooled fast reactor (LFR), may have strong positive void reactivity coefficient due to the hard neutron spectrum, which is unfavorable for reactor reactivity control. This paper presents a detailed research on the lead void effect for deeper understanding on the causes. Through analytical analysis and numerical simulation, the lead void effect was decomposed into several components, and the spatial distribution of lead void worth was also calculated. The calculation results have shown that spectral hardening effect, which will cause an increase in fertile fission (mainly 238U) and decrease in the capture rate of all heavy nuclides, accounts for the high positive lead void reactivity. Meanwhile, it has been further concluded that high positive lead void reactivity mainly exists in active region. Finally, comparison between two different lead coolant lost scenarios were carried out to further prove these conclusions.

Validation of the TWOPORFLOW code for the core analysis of liquid metal-cooled reactor with selected experiments

TWOPORFLOW (TPF) is a thermal–hydraulic code that utilizes a porous medium approach in three-dimensional Cartesian coordinates. It can simulate the flow phenomenon in fuel assemblies or reactor cores by solving three conservation equations (single-phase) or six conservation equations in a two-fluid liquid–vapor approach. Many constitutive correlations e.g. pressure loss, interfacial friction, and heat transfer (wall, liquid–vapor interface), are included to close the system of equations. The code was recently extended to simulate pre-Critical Heat Flux (CHF) and CHF heat transfer relevant for Boiling Water Reactor (BWR). In this paper, the code's simulation capability for liquid metal-cooled reactors core is validated using three experimental data sets: pressure drop of the 19-pin rod in THEADES LBE loop, the BREST-type reactor benchmark problem, and the NACIE-UP Block Fuel Pin Simulator (BFPS) test. The results show that TPF can be used not only for square rod bundle arrangements but also for hexagonal arrangements; the code can reasonably predict the thermal–hydraulic behavior of liquid metal-cooled reactors under normal and off-normal conditions.

An analytic approach to the design of passively safe lead-cooled reactors

A methodology to assist the design of liquid metal reactors, passively cooled by natural circulation during off-normal conditions, is derived from first principle physics. Based on this methodology, a preliminary design of a small LFR is accomplished and presented with accompanying neutronic and reactor dynamic characterizations. The benefit of using this methodology for reactor design compared to other available methods is discussed.

Assessment of low- and high-fidelity turbulence models for heat transfer predictions in low-prandlt number flows

Linear eddy viscosity-based Reynolds-averaged Navier-Stokes (RANS), hybrid RANS/Large Eddy Simulation (LES), and LES models are applied for simulations of plane channel flow at Reynolds number Reτ = 150 and 640. Results are obtained for fluids with Prandtl number (Pr) of 0.025 and 0.71, for cases without buoyancy (Gr = 0) and for a vertical channel flow with buoyancy (Ri = 0.1–0.13). The objective is to evaluate the predictive capability of each class of model for low-Pr flow typical of liquid metal reactor cooling systems. For this purpose, RANS, partially-averaged Navier-Stokes (PANS) and LES computations are performed on grids which contain 0.67%, 2%, and 5% of the grid points used in the Direct Numerical Simulation (DNS), respectively, and the predictions of the mean and turbulent flow and thermal quantities are validated against DNS results. RANS results obtained using Kays variable turbulent Prandtl number (PrT) formulation were more accurate than those using constant PrT for low-Pr forced convective flows, but did not play a significant role for mixed convective flows. The variable PrT formulation likewise did not have significant effect for hybrid RANS/LES or LES computations, wherein > 90% turbulence was resolved. The RANS model performed reasonably well for flows without buoyancy, but for mixed convective flows, significantly under predicted the sharp pressure gradient around the mid-channel, and the limitations were identified due to over prediction of heat transfer by wall-normal fluctuations on the aiding side. Both PANS and dynamic Smagorinsky (DSM) showed similar predictions, and under predicted the mid-channel temperature gradient for the low-Pr vertical channel case and over predicted flow turbulence on the opposing flow side and thermal fluctuations on the aiding flow side. The Wall-Adapting Local Eddy-viscosity (WALE) model performed best among the LES subgrid stress models. The improved predictions by WALE over PANS and DSM were identified because of higher modeled dissipation predictions away from the wall. Future work will focus on investigation of non-linear RANS models and extend the validation for higher Re flows.

A multi-stage approach of simulating turbulence-induced vibrations in a wire-wrapped tube bundle for fretting wear prediction

MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is a prototype of a Generation IV reactor that will be constructed in Mol, Belgium. It is a liquid metal fast reactor, using lead-bismuth eutectic (LBE) as a coolant. Flow-induced vibrations resulting in fretting wear could be of concern due to the high density of LBE. Additionally, a wire is wrapped around each fuel rod to preserve their mutual distance in the hexagonal array, which affects the flow pattern.In this research turbulence-induced vibrations are investigated. Large-eddy simulation is a sound technique to acquire turbulent loads, but it is computationally demanding to combine this in a two-way coupled fluid–structure interaction (FSI) simulation. However, because accurate prediction of the vibration of cylinders subject to liquid flow typically requires a two-way simulation due to the added-mass effect, a methodology consisting of multiple steps is proposed. First, the modal characteristics of the pin are determined in a fully coupled FSI simulation of a bare bundle. The material properties used in this model are modified to take into account the effect of the wire, which was determined experimentally beforehand. The resulting characteristics are then used to construct a realistic structural model, employed in a one-way simulation with turbulent loads from a large-eddy simulation of a wire-wrapped bundle.In this structural model contact with neighboring rods and the hexagonal duct is detected, providing realistic boundary conditions to the rod and allowing to extract contact forces. These forces are further post-processed together with the sliding displacement to calculate work rates, which is an important parameter for fretting wear prediction. Two pins of a 19-pin bundle were investigated, the central pin and a pin located in one of the corners. The simulation is repeated for multiple friction coefficients, and its impeding effect of friction on sliding motions was confirmed to moderate fretting.

Benchmark Analysis on Loss-of-Flow-without-Scram Test of FFTF Using Refined SAC-3D Models

The Fast Flux Test Facility (FFTF) is a liquid sodium-cooled nuclear reactor designed by the Westinghouse Electric Corporation for the U.S. Department of Energy. In July 1986, a series of unprotected transients were performed to demonstrate the passive safety of FFTF. Among these, a total of 13 loss-of-flow-without scram (LOFWOS) tests were conducted to confirm the liquid metal reactor safety margins, provide data for computer code validation, and demonstrate the inherent and passive safety benefits of specific design features. In our preliminary work, we have performed relatively coarse modeling of the FFTF. To better predict the transient behavior of FFTF LOFWOS test #13, we modeled it using a more refined thermal-hydraulics model. In this paper, we simulate FFTF LOFWOS test #13 with the system safety analysis code SAC-3D according to the benchmark specifications provided by Argonne National Laboratory (ANL). The simulation range includes the primary and secondary circuits. The reactor core was modeled by the built-in 3D neutronics calculation module and the parallel-channel thermal-hydraulics calculation module. To better predict the reactivity feedback introduced by coolant level variations within the GEMs, a real-time macro cross-section homogenization processing module was developed. The steady-state power distribution was calculated as the transient simulation initial boundary conditions. In general, both the steady-state calculation results and the whole-plant transient behavior predictions are in good agreement with the measured data. The relatively large deviations in transient simulation occur in the outlet temperature predictions of the PIOTA in row 6. It can be preliminarily explained by the reason for neglecting the heat transfer between channels in this model.

A New Uncertainty-Based Control Scheme of the Small Modular Dual Fluid Reactor and Its Optimization

The small modular dual fluid reactor is a novel variant of the Generation IV molten salt reactor and liquid metal fast reactor. In the primary circuit, molten salt or liquid eutectic metal (U-Pu-Cr) is employed as fuel, and liquid lead works as the coolant in the secondary circuit. To design the control system of such an advanced reactor, the uncertainties of the employed computer model and the physicochemical properties of the materials must be considered. In this paper, a one-dimensional model of a core is established based on the equivalent parameters achieved via the coupled three-dimensional model, taking into account delayed neutron precursor drifting, and a power control system is developed. The performance of the designed controllers is assessed, taking into account the model and property uncertainties. The achieved results show that the designed control system is able to maintain the stability of the system and regulate the power as expected. Among the considered uncertain parameters, the reactivity coefficients of fuel temperature have the largest influence on the performance of the control system. The most optimized configuration of the control system is delivered based on the characteristics of uncertainty propagation by using the particle swarm optimization method.

Preliminary Multi-Physics Performance Analysis and Design Evaluation of UO2 Fuel for LBE-Cooled Subcritical Reactor of China Initiative Accelerator Driven System

The China initiative Accelerator Driven System (CiADS) and the corresponding lead-bismuth eutectic (LBE) cooled subcritical reactor, as the research subject of one of the major national science and technology infrastructure projects, are undertaken by the Institute of Modern Physics-Chinese Academy of Sciences (IMP-CAS). And in the first phase, UO2 fuels will be loaded in the subcritical core to test the coupling technology and achieve a long-term steady operation. A brief description of CiADS subcritical reactor, fuel assembly and fuel element are presented here, and a multi-physics performance analysis and design evaluation of CiADS UO2 fuel are carried out by means of the FUTURE code. FUTURE is a fuel performance analysis code to evaluate the synergy of phenomena occurring in the fuel element and their impact on the fuel design improvement for the liquid metal fast reactor, which was developed jointly by IMP-CAS and Xi’an Jiaotong University (XJTU). In this paper, the FUTURE code was modified and updated focusing on characteristics of CiADS fuels. Relocation and densification models were added. Results of the hottest fuel element, mainly concerning the thermo-mechanical behaviors, are discussed concerning both fuel and cladding performance on the basis of indicative design limits. According to the preliminary design, the CiADS UO2 fuel exhibits good performance, and the main safety parameters are far below the indicative limits. The Fuel Cladding Mechanical Interaction (FCMI) is not very serious, and the permanent cladding strains and Cumulative Damage Fraction (CDF) are small and even negligible thanks to the low level of fuel temperature and corresponding stress. However, some critical issues may still exist, especially on LBE corrosion near the coolant inlet, where protective oxide layers are very thin from BoL to EoL. The modeling is useful for providing feedback to the conceptual design of the CiADS LBE-cooled subcritical reactor and the update of FUTURE code.

Post-irradiation examination of optical components for advanced fission reactor instrumentation.

The use of optical instrumentation in advanced nuclear fission systems, such as molten salt reactors, liquid metal-cooled reactors, and high-temperature gas-cooled reactors, has the potential to enhance reactor safety and economic performance through in situ and online measurement of reactor conditions. Selection of suitable optical components, such as optical windows and fibers, is essential for operation of optical instrumentation in intense radioactive and thermal environments inherent to nuclear reactor systems. We present the development and performance of a self-contained and mobile post-irradiation examination system for rapid characterization of the optical properties of materials. The instrument combines linear absorption and nanosecond Z-scan modules in a compact, relocatable design. The system mobility allows for the evaluation of optical samples at the site of irradiation, minimizing the delay between extraction from the irradiation site and optical characterization. This provides nearly real-time information on the material performance under simultaneous irradiation and thermal annealing, simulating the relevant conditions for the use of those components in nuclear power systems.