Liquid Metal Fast Reactor Research Articles

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.

Numerical Simulation of Turbulent Mixing of LBE between Sub-Channels of Wire-Wrapped Fuel Assembly

The lead-bismuth eutectic (LBE) is a candidate coolant for the fourth generation liquid metal reactors. In view of its special thermophysical properties, its flow and heat transfer process in the fuel assembly sub-channels need to be studied. For this purpose, the authors conduct numerical simulation and analysis of the turbulent flow of the LBE in the wire-wrapped fuel assembly, and compare the numerical simulation results of the wall temperature of fuel rods with the experimental data for response. The simulation results agree well with the experimental data, indicating that the mathematical model and numerical results are highly reliable and accurate. The authors also characterize the turbulent mixing of LBE between different sub-channels at different values of the ratio of spacing between fuel rods to the fuel rod diameter (S/D), using the turbulent mixing coefficient β. The characterization results show that the fluctuation of β between different sub-channels varies and that the β value is negatively correlated with the S/D. Based on different calculation results of the S/D and Renolyd number, the correlation formula for the β between different sub-channels is obtained by fit. It provides a turbulent mixing model for the development of a sub-channel analysis code of fuel assembly consisting of wire-wrapped fuel rods arranged in triangular form.

Neutronics modelling of control rod compensation operation in small modular fast reactor using OpenMC

The small modular liquid-metal fast reactor (SMFR) is an important component of advanced nuclear systems. SMFRs exhibit relatively low breeding capability and constraint space for control rod installation. Consequently, control rods are deeply inserted at beginning and are withdrawn gradually to compensate for large burnup reactivity loss in a long lifetime. This paper is committed to investigating the impact of control rod compensation operation on core neutronics characteristics. This paper presents a whole core fine depletion model of long lifetime SMFR using OpenMC and the influence of depletion chains is verified. Three control rod position schemes to simulate the compensation process are compared. The results show that the fine simulation of the control rod compensation process impacts significantly the fuel burnup distribution and absorber consumption. A control rod equivalent position scheme proposed in this work is an optimal option in the trade-off between computation time and accuracy. The control position is crucial for accurate power distribution and void feedback coefficients in SMFRs. The results in this paper also show that the pin level power distribution is important due to the heterogeneous distribution in SMFRs. The fuel burnup distribution at the end of core life impacts the worth of control rods.

Electromagnetic Pump ForLarge Pool Concept Liquid Metal Fast Breeder Reactor

Electromagnetic Pumps are passive pump system and have been integrated in use for pumping liquid sodium in auxiliary cooling circuits such as fill and drain from the container of such metallic liquid, and purification circuits of sodium cooled fast breeder nuclear reactors, such as Liquid Metal Fast Breeder Reactor

In situ Monitoring of the Oxidation of Coated and Uncoated Austenitic Stainless Steels at Temperatures Lower than 600 °C: Part A—Coating Screening

The lithium niobate used in an ultrasonic transducer for liquid metal fast nuclear reactors is prone to a reduction leading to a rapid short circuit of the sensor in operating conditions. Oxygen consumption on stainless steel sensor’s walls was suspected to be the cause of the reduction. Impedance spectroscopy monitoring around 600 °C under different O2 partial pressures shows that the reduction limit is above the equilibrium pressure of iron oxides. As the housing material (304L) cannot be easily changed, various coating solutions enhancing the passivation of the steels were investigated. Sol–gel lanthanum coatings show the best behavior when tested by TGA at 800 °C. However, in situ monitoring of the oxygen consumption by an oxygen sensor close to operating condition shows that it was not enough protective.

RANS simulations for transition and turbulent flow regimes in wire-wrapped rod bundles

Internal flows through rod assemblies are commonly found in heat exchangers, steam generators, and nuclear reactors. One of the fuel assembly designs considered for liquid metal-cooled reactors utilizes wires helically wrapped around each fuel rod as spacers. The wires keep the fuel pins separated, enhancing the turbulent mixing, and heat transfer, but also affecting the pressure drop. It is of interest the understanding of the fluid flow phenomena in the sodium-cooled fast reactor as it is one of the Generation IV advanced reactor designs and it has been a motivating topic of research for the last decade. A wire-wrapped fuel assembly replica with 61-pins has been in operation at the Thermal-Hydraulic Research Laboratory of Texas A&M University. This facility produced high-fidelity velocity and pressure drop data for validation of computational fluid dynamics codes. This study investigates the effects of geometrical features and operating conditions on the flow behavior of the 61-pin wire wrapped bundle using Reynolds-Averaged Navier-Stokes (RANS) models to predict the axial and transverse pressure drops for a range of Reynolds numbers from 1,270 to 100,000. The friction factor predictions were in satisfactory agreement with the experimental data and the Upgraded Chen and Todreas correlation. The internal subchannel velocity results were compared with experimental data and Large Eddy Simulations (LES) and found in reasonable agreement. This study demonstrates that RANS is a suitable approach in predicting velocity and pressure fields in wire-wrapped rod bundles, with a relatively low computational effort.

Corrosion Behavior and Surface Treatment of Cladding Materials Used in High-Temperature Lead-Bismuth Eutectic Alloy: A Review

Liquid metal fast reactors were considered to be the most promising solution to meet the enormous energy demand in the future. However, corrosion phenomenon caused by the liquid metal, especially in high-temperature lead-bismuth coolant, has greatly hindered the commercialization of the advanced Generation-IV nuclear system. This review discussed current research on the corrosion resistance of structural materials (such as EP823, T91, ODS, and authentic steels) in high-temperature liquid metal served as reactor coolants. The current corrosion resistance evaluation has proved that even for the excellent performance of EP823, the structural material selected in pressurized water reactor is not the ideal material for operation in the high-temperature lead-bismuth eutectic (LBE). Furthermore, the latest coating technologies that are expected to be applied to cladding materials for coolant system were extensively discussed, including Al-containing coatings, ceramic coatings, oxide coatings, amorphous coatings and high-entropy alloy coatings. The detailed comparison summarized the corrosion morphology and corrosion products of various coatings in LBE. This review not only provided a systematic understanding of the corrosion phenomena, but also demonstrated that coating technology is an effective method to solve the corrosion issues of the advanced next-generation reactors.

Cardinal: A Lower-Length-Scale Multiphysics Simulator for Pebble-Bed Reactors

This paper demonstrates a multiphysics solver for pebble-bed reactors, in particular, for Berkeley’s pebble-bed -fluoride-salt-cooled high-temperature reactor (PB-FHR) (Mark I design). The FHR is a class of advanced nuclear reactors that combines the robust coated particle fuel form from high-temperature gas-cooled reactors, the direct reactor auxiliary cooling system passive decay removal of liquid-metal fast reactors, and the transparent, high-volumetric heat capacitance liquid-fluoride salt working fluids (e.g., FLiBe) from molten salt reactors. This fuel and coolant combination enables FHRs to operate in a high-temperature, low-pressure design space that has beneficial safety and economic implications. The PB-FHR relies on a pebble-bed approach, and pebble-bed reactors are, in a sense, the poster child for multiscale analysis. Relying heavily on the MultiApp capability of the Multiphysics Object-Oriented Simulation Environment (MOOSE), we have developed Cardinal, a new platform for lower-length-scale simulation of pebble-bed cores. The lower-length-scale simulator comprises three physics: neutronics (OpenMC), thermal fluids (Nek5000/NekRS), and fuel performance (BISON). Cardinal tightly couples all three physics and leverages advances in MOOSE, such as the MultiApp system and the concept of MOOSE-wrapped applications. Moreover, Cardinal can utilize graphics processing units for accelerating solutions. In this paper, we discuss the development of Cardinal and the verification and validation and demonstration simulations.

Wetting property of liquid Pb on different steel candidates as structural materials for the Generation IV nuclear reactors

Conventional high-chromium ferritic / martensitic steels, such as T91 and AISI 316 L as well as the new oxide dispersed strengthened (ODS) steels, i.e. Fe-Cr (ferritic or martensitic) with a fine Y2O3 dispersion, are promising candidates as structural materials for applications in aggressive environments. In order to improve the efficiency and lifetime of structural materials for liquid metal fast reactor (LMFR), where liquid lead (Pb) is used as coolant, the investigations are currently focused on the interfacial phenomena occurring at the liquid Pb/steel interfaces. The better understanding of the interfacial phenomena is the key issue for the selection of structural materials meeting the requirements of the demanding application. In the present work, a fundamental study on the wetting characteristics and the interfacial phenomena of liquid Pb in contact with T91, AISI 316L, ODS- 12Cr and ODS-14Cr steels over the temperature range 550�820�C and under different oxygen containing atmospheres is reported.

Numerical Study on the Thermal Hydraulic Characteristics in a Wire-Wrapped Assembly of LFRs

The concept of wire-wrapped assembly has been widely adopted in liquid metal-cooled reactors to enhance the capability of flow and heat transfer. The detailed 3D thermal hydraulic behaviors of liquid metal in the wire-wrapped assembly directly affect the safety of the reactor. In the present work, a numerical study on a 19-pin wire-wrapped assembly cooled by a lead-bismuth eutectic was carried out using the CFD method. The geometrical model is referred to in the test section in KIT’s 19-pin experiments, which is a scaled test assembly based on the MYRRHA reactor. The established CFD model was validated against experimental data and results showed that the SST k-ω model accompanied with the modified Prt was appropriate for the current problem. Then the numerical models were extensively used in the modeling of the blocked assembly flow with a cylindrical blockage. Also, the influences of blockage longitude, position, and porosity on the thermal hydraulic characteristics in the assembly were evaluated in detail. This work is meaningful for the deep understanding of safety features under the condition of assembly blockage operation in LFRs.

The Development of the Technology for Producing Vessel Structures of the Main Circulation Pump for a BREST Reactor

Abstract—One of the joints of a liquid metal fast neutron reactor is the main circulation pump. The fabrication technology of the most labor-intensive joint of the main circulation pump, the pressure chamber, which is a complex welded structure made of 10Kh15N9S3B3 (EP-302Sh) sheet steel, is optimized and the fixtures and accessories are developed and fabricated, as a result of which the metal content of the structure is decreased. The process documentation for bending and stamping of the parts of the pressure chamber and welding of the components and their thermal and mechanical treatment and assembling is developed. The existing process documentation is adjusted as applied to the pressure chamber of the main circulation pump following the results of fabrication of this structure.

Numerical simulation on LMR molten-core centralized sloshing benchmark experiment using multi-phase smoothed particle hydrodynamics

The Smoothed Particle Hydrodynamics is one of the most widely used mesh-free numerical method for thermo-fluid dynamics. Due to its Lagrangian nature and simplicity, it is recently gaining popularity in simulating complex physics with large deformations. In this study, the 3D single/two-phase numerical simulations are performed on the Liquid Metal Reactor (LMR) centralized sloshing benchmark experiment using the SPH parallelized using a GPU. In order to capture multi-phase flows with a large density ratio more effectively, the original SPH density and continuity equations are re-formulated in terms of the normalized-density. Based upon this approach, maximum sloshing height and arrival time in various experimental cases are calculated by using both single-phase and multi-phase SPH framework and the results are compared with the benchmark results. Overall, the results of SPH simulations show excellent agreement with all the benchmark experiments both in qualitative and quantitative manners. According to the sensitivity study of the particle-size, the prediction accuracy is gradually increasing with decreasing the particle-size leading to a higher resolution. In addition, it is found that the multi-phase SPH model considering both liquid and air provides a better prediction on the experimental results and the reality.

Experimental Characterization of Leak Detection Systems in HLM Pool Using LIFUS5/Mod3 Facility

In the framework of the European Union MAXSIMA project, the safety of the steam generator (SG) adopted in the primary loop of the Heavy Liquid Metal Fast Reactor has been studied investigating the consequences and damage propagation of a SG tube rupture event and characterizing leak rates from typical cracks. Instrumentation able to promptly detect the presence of a crack in the SG tubes may be used to prevent its further propagation, which would lead to a full rupture of the tube. Application of the leak-before-break concept is relevant for improving the safety of a reactor system and decreasing the probability of a pipe break event. In this framework, a new experimental campaign (Test Series C) has been carried out in the LIFUS5/Mod3 facility, installed at ENEA Centro Ricerche Brasimone, in order to characterize and to correlate the leak rate through typical cracks occurring in the pressurized tubes with signals detected by proper transducers. Test C1.3_60 was executed injecting water at about 20 bars and 200°C into lead-bismuth eutectic alloy. The injection was performed through a laser microholed plate 60 μm in diameter. Analysis of the thermohydraulic data permitted characterization of the leakage through typical cracks that can occur in the pressurized tubes of the SG. Analysis of the data acquired by microphones and accelerometers highlighted that it is possible to correlate the signals to the leakage and the rate of release.

Wall resolved large eddy simulation of reactor core flows with the spectral element method

Numerical simulation is an intrinsic part of nuclear engineering research supporting the design of nuclear power plants. With the advent of Petascale computing (i.e., computers capable of more than 1 PFlop), the simulation of portions of reactor components with turbulence-resolving techniques is now possible. These simulations can provide invaluable insight into the flow dynamics, which is difficult or often impossible to obtain with experiments alone. The spectral element method in particular has emerged as a powerful method to deliver massively parallel calculations at high fidelity by using Large Eddy Simulation or Direct Numerical Simulation. In this work we review the fundamentals of the method and the reasons it is compelling for the simulation of nuclear reactor core flows. We review a series of wall-resolved large eddy simulations of reactor cores applied to three reactor types. In particular we discuss the simulation of the flow in a 17 × 17 rod bundle related to light water reactor (LWR) applications; the flow in a 37-pin wire-wrapped rod bundle related to liquid metal reactor (LMR) applications and the flow through a graphite block with application to molten salt reactors (MSR).

ФУНДАМЕНТАЛЬНЫЕ ИССЛЕДОВАНИЯ В ОБЛАСТИ ТЕПЛОФИЗИКИ ЯЭУ: ДОСТИГНУТЫЕ РЕЗУЛЬТАТЫ И ПРОБЛЕМЫ ДАЛЬНЕЙШИХ ИССЛЕДОВАНИЙ

The results of experimental and theoretical calculations of fundamental research in the field of hydrodynamics, heat transfer, physical chemistry and nuclear power technology cooled by water and liquid metal coolants are analyzed, problems and tasks of further research are formulated. The effects and physical phenomena, laws and characteristics for a wide range of processes occurring in the first and second circuits of reactor facilities, including the flow part, the core, the heating equipment, the means of monitoring and cleaning coolants from impurities as applied to watercooled thermal neutron reactors, fast liquid metal reactors, fusion plants, etc. Particular attention in the field of hydrodynamics and heat transfer is given to studies of the heat transfer crisis in watercooled reactors, hydrodynamics of collector systems, stratification and mixing of jets, vibroacoustics, turbulent transfer in complex channels, heat transfer in the channels and assemblies of fast reactor fuel elements, contact thermal resistance, boiling in large volume and bundles of fuel rods, condensation of liquid metals. In the field of physical chemistry and the technology of liquid metal coolants, the types of impurities and the sources of their entry into the coolants for various nuclear power plants, the characteristics of the mass and heat transfer processes in the circuits and equipment of reactor facilities with liquid metal coolants (sodium, potassium, lithium, lead, bismuth lead), the efficiency of using various purification devices (cold and hot traps) and monitoring the state of impurities in coolants are analyzed.Problem questions and suggestions for further research are formulated. Information is given on the key thermophysical problems of basic research in relation to the development of innovative nuclear energy technologies: water-cooled reactors with supercritical water pressure, high-temperature fast reactors with sodium coolant for hydrogen production, thermonuclear installations, nuclear power plants with liquid-metal space coolants.

Technology for producing vessel structures of the main circulation pump for BREST reactor

One of the important joints of a liquid metal fast reactor is the main circulation pump. In the course of the present research, a preproduction model of the pressure chamber, the most labor-consuming unit of the main circulation pump, was manufactured. This complex welded structure was made of 10Kh15N9S3B3 (EP-302Sh) steel sheets; fixtures and accessories were also manufactured, and as a result the metal consumption of the structure decreased. Technological documentation was made for bending, stamping of parts of the pressure chamber, welding of components, their heat and mechanic treatment, and its assembling. According to the results of this work, the existing technological documentation was adjusted in the context of the new data for pressure chamber of the main pump.

Experimental validation of simulating natural circulation of liquid metal using water

Liquid metal-cooled reactors use various passive safety systems driven by natural circulation. Investigating these safety systems experimentally is more advantageous by using a simulant. Although numerous experimental approaches have been applied to natural circulation-driven passive safety systems using simulants, there has been no clear validation of the similarity law. To validate the similarity law experimentally, SINCRO-V experiment was conducted using Wood's metal and water for simulant of the Wood's metal. A pair of SINCRO-V facilities with length-scale ratio of 14.1:1 for identical Bo’ was investigated, which was the main similarity parameter in temperature field simulation. In the experimental range of 0.2–1.0% of decay heat, the temperature distribution characteristics of the small water facility were very similar to that of the large Wood's metal facility. The temperature of the Wood's metal predicted by the water experiment showed good agreement with the actual Wood's metal temperature. Despite some error factors like discordance of Gr’ and property change along the temperature, the water experiment predicted the Wood's metal temperature with an error of 27%. The validity of the similarity law was confirmed by the SINCRO-V experiments.

Liquid metal thermal hydraulics: State-of-the-art and future perspectives

The energy consumption worldwide will increase for the coming decades. In order to address global warming, international experts agree that no low CO2 emitting energy production routes should be omitted. In fact, nuclear energy production is appointed an important role in the transition towards a low CO2 emitting world. Taking into account the identified reserves of uranium and assuming a stable consumption of uranium in light water reactors, the uranium reserves will surely fit for one or two centuries. However, by switching to another fuel cycle using uranium in fast reactors, the uranium reserves may serve the world for a multi-fold of centuries. Nowadays, the most mature fast reactor designs are liquid metal cooled fast reactors.Design support and safety analysis of liquid metal reactors strongly relies on thermal hydraulic simulations and experiments. This paper will discuss the current state-of-the-art in liquid metal thermal hydraulics and also its future perspectives. In particular, the following topics will be discussed: liquid metal heat transfer, fluid-structure interaction, fuel assembly thermal hydraulics, coolant solidification, efficient 3-d system modelling, and validation at large. For each topic, the state-of-the-art, challenges and the future perspectives will be described from the author’s point of view.

Numerical investigation on LBE-water interaction for heavy liquid metal cooled fast reactors

Pool type Reactor Pressure Vessels (RPVs) containing steam generators with high secondary coolant pressure are adopted for a number of heavy liquid metal reactor conceptual designs. The primary side heavy liquid metal, e.g., liquid Lead-Bismuth Eutectic (LBE) with low pressure and secondary water with high pressure can come into contact in the event of Steam Generator Tube Rupture (SGTR) scenarios. Due to the pressure wave propagation and resultant sloshing of liquid LBE, the vigorous interaction between water and melt LBE can threaten the structural integrity of the reactor internals. Moreover, Reactor Pressure Vessel (RPV) pressurization, reactivity feedback as a result of steam entrance into the core as well as primary system pollution are also key concerns for SGTR accidents.The article aims to evaluate the capability of the MC3D code for LBE–water interaction in the event of SGTR scenarios. MC3D is a multidimensional numerical tool developed by IRSN for multi-phase flow studies and assessments in nuclear safety. Its field of application is mainly on Fuel-Coolant Interaction (FCI), especially the stages of premix and explosion in case of severe accidents. In order to evaluate the capability of MC3D for the Coolant-Coolant Interaction (CCI) of Liquid metal cooled Fast Reactors (LFRs), simulation of a small-scale experiment investigating water jet plunging into a pool of molten lead-bismuth alloy is carried out. Afterwards, experiments on LIFUS5/Mod2 facility at ENEA studying LBE-water interaction in the event of SGTR, for which a series of experiments have been conducted including the injection of sub-cooled water into LBE from the bottom of the test vessel, is simulated with MC3D as well. The simulation results for both facilities are then compared with those of the tests. The results predicted with MC3D for LIFUS5/Mod2 is also compared against the simulation results with SIMMER. The comparison impliesthat MC3D is generally capable for the phenomena of CCI in the event of SGTR for LFRs. Nevertheless, the potential employment of MC3D for intensive investigation of the LBE-water interaction will still need further experimental studies and improvement of physical models.

Seismic and stress qualification of LMFR fuel rod and simple method for the determination of LBE added mass effect

In this study, two different designs of liquid metal fast reactor (LMFR) fuel rods wire-wrapped and non-wire-wrapped (bare) are compared with respect to different parameters as a means of considering the optimum fuel design. Nuclear seismic rules require that systems and components that are important for safety must be capable of bearing earthquake effects, and that their integrity and functionality should be guaranteed. Mode shapes, natural frequencies, stresses on cladding, and seismic aspects are considered for comparison using ANSYS. Modal analysis is compared in a vacuum and in lead–bismuth eutectic (LBE) using potential flow theory by considering the added mass effect. A simple and accurate approach is suggested for the determination of the LBE added mass effect and is verified by a manually calculated added mass, which further proved the usefulness of potential flow theory for the accurate estimation of the added mass effect. The verification of the hydrodynamic function (τ) over the entire frequency range further validated the finite element method (FEM) modal analysis results. Stresses obtained for fuel rods against different loading combinations revealed that they were within the allowable limits with maximum stress ratios of 0.25 (bare) and 0.74 (wire-wrapped). In order to verify the structural integrity of cladding tubes, stresses along the cladding length were determined during different transients and were also calculated manually for static pressure. The manual calculations could be roughly compared with the ANSYS results, and the two showed a close agreement. Contact analysis methodology was selected, and the most appropriate analysis options were suggested for establishing contact between the wire and cladding for the wire-wrapped design grid independence analysis, which proved the accuracy of the results, confirmed the selection of the appropriate procedure, and validated the use of the ANSYS mechanical APDL code for LMFR fuel rod analysis. The results provided detailed insight into the structural design of LMFR fuel rods by considering different structural configurations (i.e., bare and wire-wrapped) in the seismic loading; this not only provides a FEM procedure for LMFR fuel with complex configuration, but also guides the reference design of LMFR fuel rods.