Liquid Metal Fast Reactor Research Articles

Development of Mechanistic Source Term Analysis Tool SAS4A-FATE for Lead- and Sodium-Cooled Fast Reactors

Fauske & Associates, LLC (FAI), Argonne National Laboratory (ANL), and Westinghouse Electric Company are collaborating within the program “Development of an Integrated Mechanistic Source Term Assessment Capability for Lead- and Sodium-Cooled Fast Reactors.” This program, partially funded by the U.S. Department of Energy through the Gateway for Accelerated Innovation in Nuclear initiative, aims at developing a computational framework for predicting radionuclide release from a broad spectrum of accidents that can be postulated to occur at liquid metal cooled reactor (LMR) facilities. Specifically, the program couples the transient and severe accident analysis capability of the SAS4A/SASSYS-1 code developed by ANL with the radionuclide transport analysis capability of the Facility Flow, Aerosol, Thermal, and Explosion (FATE) code developed by FAI. The testing of both the individual codes and of the coupled system is performed on a generic lead cooled fast reactor (LFR) design that is intended to capture the key differences between the LFR and the sodium fast reactor (SFR), around which the SAS4A/SASSYS-1 code has historically been developed and from which the coupled code inherits some features requiring modification before application to LFR systems. By means of this approach, a computational framework applicable to both LFR and SFR systems will be obtained that will assist LMR developers in performing a realistic, scenario-dependent mechanistic source term (MST) assessment expected not only to strengthen their safety case but also to support easier siting and claims on reduced emergency planning zone requirements. This paper discusses the work being performed to adapt the SAS4A/SASSYS-1 and FATE codes to LFR technology; the code coupling method implemented; and some of the results of the LFR test case, with the latter aimed at demonstrating the progress made toward the development of the MST analysis capability that is ultimately targeted.

Analysis of beryllium poisoning effect on liquid metal reactor with U–Be alloy fuel

A liquid metal reactor (LMR) loaded with a fuel compound of uranium and beryllium (U–Be alloy fuel), which was cooled by a lead–bismuth eutectic alloy (PbBi), has been applied in Russian Alfa-class nuclear submarines. Because of the large amount of beryllium in the core, the reaction between the beryllium atoms and neutrons could result in the accumulation of 3He and 6Li, which are called the “poisoned elements” owing to their large thermal neutron capture cross section. The accumulation of neutron absorber can affect the performance of a reactor. In this study, the Super Multi-functional Calculation Program (SuperMC) code, which was developed by Institute of Nuclear Energy Safety Technology of the Chinese Academy of Sciences (INEST, CAS), was adopted to illustrate the influence of beryllium on an LMR.

Structural seismic characteristics assessment of LMFR fuel assembly hexagonal wrapper over operational temperature range

In this study, the structural integrity of liquid metal fast reactor fuel assembly has been established for different parameters considering the optimum fuel design. Analytical calculation of added mass effect due to lead bismuth eutectic and verification through previously presented theories, has been established. The integrity of the hexagonal wrapper of fuel assembly has been guaranteed over the entire operating temperature range. Effect of temperature on the density of lead bismuth eutectic, the subsequent change in added mass of lead bismuth eutectic, the effect on natural frequencies and effect on stresses on wrapper, has been studied in detail. A simple empirical relationship is presented for estimation of added mass effect for lead bismuth eutectic type fast reactors for any desired temperature. An approach for assessment of fast reactor fuel assembly performance has been outlined and calculated results are presented. Nuclear seismic rules require that systems and components which are important to safety, shall be capable of bearing earthquake effects and their integrity and functionality should be guaranteed. Mode shapes, natural frequencies, stresses on wrapper and seismic aspect has also been considered using ANSYS. Modal analysis has been compared in vacuum and lead bismuth eutectic using the calculated added mass.

CFD ANALYSIS OF FLOW BLOCKAGE IN THE ALFRED FA

In the context of GEN-IV heavy liquid metal-cooled reactors safety studies, the flow blockage in a fuel subassembly is considered one of the main issues to be addressed and one of the most important and realistic accident for Lead Fast Reactors (LFR) fuel assembly. The blockage in a fast reactor Fuel Assembly (FA) may have serious effects on the safety of the plant leading to the FA damaging or melting. The external or internal blockage of the FA may impair the correct cooling of the fuel pins, be the root cause of anomalous heating of the cladding and of the wrapper and potentially impact also fuel pins not directly located above or around the blocked area. In order to model the temperature and velocity field inside a wrapped FA under unblocked and blocked conditions, detailed CFD thermal hydraulic analyses of the FA are required. ALFRED is a 300 MWth Lead-cooled fast reactor GEN.IV concept. The advanced conceptual design of the reactor was carried out within the LEADER EU project in the last years. In the SESAME EU project specific thermal-hydraulic experiments and simulations were addressed to explore the basic phenomenology of some accidental events like internal flow blockage. A CFD computational model of the ALFRED FA is built for thermal hydraulic analysis of relevant internal blockage scenarios. The whole model of the ALFRED Fuel Assembly is first presented and calculation of flow and temperature field in nominal condition is carried out. RANS simulations of idealized blockage scenarios like one sector and two-sectors blockage are performed adopting three different spacer grid location (under the active length, at half active length, above the active length). Results showed that the most likely blockage in the lower grid positioned before the active region do not perturb the temperature distribution in the fuel assembly, while the central grid ones have strong consequences and leads to a clad temperature peak behind the blockage with possible clad failure. On the other hand, the most severe case on the upper grid (2 sectors blockage) showed a manageable temperature maximum (700°C) at the end of the active region due to the lower mass flow rate in the FA. As a conclusion lower and upper grid blockages perturb only marginally the temperature field and do not lead to serious clad failure involving only creep long-term effects, while a blockage in the active region leads to serious clad damage and possible clad deformation and fusion.

Performance evaluation of eddy current flowmeter in Monju

ABSTRACT Measurement of the temperature and flow rate at each fuel subassembly outlet is an effective way for a liquid metal fast breeder reactor to detect a loss of coolant accident or reactivity-initiated accident in the early stage and to understand the reactor’s thermal hydrodynamic performance. Japan Atomic Energy Agency has developed the eddy current flowmeter in practical use and installed 34 of them in the upper core structure of fast breeder reactor, Monju. This report presents data obtained by using the flowmeters in Monju. We observed high linearity between each of the flowmeter’s signal intensity and the primary sodium’s flow rate under 10–100% flow rate condition. High linearity was also observed in a region of low velocity (approx. 0.25 m/s). The fluctuation of flow rate observed by the flowmeters was below 0.2 m/s which is 5% of the time-averaged velocity under a rated condition. These experimental results show that the eddy current flowmeter is an effective tool to detect the changes in relative flow rate.

Experimental Simulation of Hydrodynamics and Heat Transfer in Bubble and Slug Flow Regimes in a Heavy Liquid Metal

For the confirmation of the claimed design properties of a reactor plant with a heavy liquid-metal coolant, computational and theoretical studies should be performed in order to justify its safety. As one of the basic scenarios of an accident, the leakage of water into the liquid metal is considered in the case of steam generator tube decompression. The most important in the analysis of such a kind of accidents are questions concerning the motion and heat exchange of steam bubbles in the steam generator and the probability of blocking the flow area owing to freezing coolant, since the temperature of boiling feedwater in the steam generator can become lower than the melting point of lead. In this paper, we present main approaches and relationships used for the simulation of the motion of gas bubbles and heat transfer between bubbles and liquid metal flow. A brief description of the HYDRA-IBRAE/LM computational code that can be used to analyze emergency situations in a liquid metal-cooled reactor facility is also presented. It should be noted that the existing experimental data on the motion and heat transfer of gas bubbles in a heavy liquid metal are insufficient. For this reason, in order to verify the HYDRA-IBRAE/LM code models, experiments have been performed on the cooling of liquid lead by argon and on the motion of gas bubbles in the Rose’s alloy. In particular, a change in the temperature of the coolant over time has been studied, and the void fraction of gas at different flow rates of gas has been measured. A detailed description of the experiments and a comparison of the results of the calculations with the experimental data are presented. The analysis of uncertainties made it possible to reveal the main factors that exert the greatest effect on the results of calculations. The numerical analysis has shown that the models incorporated into the HYDRA-IBRAE/LM code allow one to describe to a sufficient degree of confidence the process of cooling liquid lead melt when argon bubbles pass through it, simulating the flow of water into the liquid metal in the course of steam generator tube rupture.

High fidelity numerical simulations of an infinite wire-wrapped fuel assembly

Abstract Liquid metal fast reactors (LMFRs) are foreseen to play a relevant role in the future of nuclear energy, thanks to their increased fuel utilization, high power density and intrinsic safety deriving from the optimal heat transfer performance of the metallic coolants. As these concepts adopt very tight fuel packing, accurate thermal-hydraulic analysis of their fuel assemblies is recognized as a crucial scientific and engineering contribution in order to support the deployment of such technology, and has challenged the modeling and simulation community for several decades. The fuel pins in most liquid metal cooled reactor assemblies are separated by wire wrap spacers, which are helically wrapped wires around the pin along its axis. The wire wraps were first introduced to maintain a constant spacing among the fuel pins, but further introduce additional turbulence and establish a bulk rotation of the fluid within the assembly. The flow around the pins is fully turbulent, with Reynolds number in the range of 40,000–65,000. Accurate local predictions of coolant flow and heat transport with Computational Fluid Dynamics techniques in such a complex flow configuration stress both the accuracy of available turbulence models and the availability of computational power. Additionally, only incomplete validation of the modelling approaches is available, especially with respect to heat transport, due to the very limited number of experimental databases. The aim of the present study has been to generate a high fidelity reference database for a wire-wrapped fuel assembly for application to LMFR fuel. The database is openly available to the scientific community and can serve as a reference to validate the low order turbulence modelling approaches, such as hybrid (URANS/LES) methods and RANS. The dimensions of the geometrical configuration are based on the Multi-purpose hybrid research reactor for high-tech applications, MYRRHA, fuel design. Extensive analyses of flow and thermal fields have been performed and are discussed in this article.

Demonstration of Under Sodium Viewer in Monju

Development of an inspection technique in opaque liquid-metal coolant is one of the important issues to ensure the safety of the liquid-metal fast breeder reactor (LMFBR). Performance tests of an under sodium viewer (USV), which was developed to detect an obstacle in the reactor vessel (RV) of the LMFBR Monju, have been carried out. Ultrasonic sensors and reflectors are located across the core inside Monju’s RV. The USV can detect an obstacle existing between the core top and the upper core structure bottom by identifying differences of echo signals. This paper describes the USV performance tests. In the tests, the reference echo signals under various conditions were accumulated, and the signal-to-noise ratio successfully exceeded the target value. Measured signals clearly differed with and without an obstacle. These experimental results show the performance of the USV for detecting an obstacle in a specified place.

Vibration of core subassemblies due to large sodium–water reaction in the steam generator of a Liquid Metal Fast Breeder Reactor

In a Liquid Metal cooled Fast Breeder Reactor (LMFBR), liquid sodium is used as the reactor coolant to extract the heat energy from the core of the reactor. Liquid sodium has a chemical characteristic that makes it to violently react with water or steam. When large sodium-water reaction occurs due to defect, leakage or breakage in the steam generator tubes, high pressure waves will be generated and get transmitted throughout the secondary sodium circuit and interact with the components like sodium headers on top/bottom of steam generator, surge tank, intermediate heat exchanger etc., which also affects the safety of primary sodium circuit and core subassemblies. In LMFBR, sodium water reactions are included in the Design Basis Events. This dynamic event needs to be analyzed from the perspective of integrity of core subassemblies considering the phenomenon of fluid structure interaction. To evaluate the integrity of core subassemblies, the behavior of high pressure waves at intermediate heat exchanger is considered and the response of the subassemblies to the dynamic pressure load has been studied. The dynamic analysis of the reactor model is carried out using a three-dimensional finite element analysis in CAST3M computer code. In the code, the fluid elements are represented by volume elements and the coupling between fluid and structure by surface elements. Fluid damping is considered in the simulation. Validation of the code has been carried out using an experimental setup. Analysis results shows the maximum displacement of the primary circuit components viz., IHX, Inner vessel and subassemblies are 5.0 mm, 2.0 mm and 0.07 mm respectively for a design basis sodium water reaction event. Due to this displacement there is negligible change in the reactivity of the core.

Assessment study about the risk of Wilks’ formula for uncertainty quantification of design extension condition scenarios in prototype Gen-IV sodium fast reactor

ABSTRACT Sensitivity analysis and uncertainty quantification using Wilks’ formula and Monte Carlo for Unprotected Loss of Flow (ULOF) and Unprotected Transient OverPower (UTOP) accidents of prototype Gen-IV sodium-cooled fast reactor were performed. Multi-dimensional analysis for reactor safety for liquid metal reactors code calculations were conducted while simultaneously varying the values of all uncertain parameters according to their distribution using parallel computing platform integrated for uncertainty and sensitivity analysis to obtain uncertainty bands for Figures of Merit (FOM) – coolant, fuel centerline, and cladding temperature at the hottest fuel rod. To specify the uncertainty range of the parameters for each accident scenario, literature survey and expert judgments were consulted. By the sensitivity analysis, the importance ranking of 25 parameters in model identification and ranking table based on phenomena identification and ranking table was identified. Through Monte Carlo calculation, 95% upper limit and 95% confidence level were obtained, and about 2% and 5% under-prediction (risk) of FOM of ULOF and UTOP accidents using Wilks’ formula were confirmed, respectively.

Evaluation on Creep Properties of Type 316SS Series

Evaluation on creep properties of type 316 stainless steel (SS) series has been done. Type of 316SS series is a candidate material for major structural components of high temperature reactor such as liquid metal reactor (LMR) because it has a good mechanical property in high temperature, compatibility with the coolant sodium, and adequate welding ability. The objective of this research is to obtain and compare creep properties between type 316SS and type 316LN SS materials. The method use by conducted creep rupture tests in ranges temperature of 600°C, 700°C and 800°C and under a constant applied stress level of 200 MPa (393 N). Research result obtained that time to rupture of type 316SS less than that of type 316LN SS. The values of extension (mm), creep strain (mm/mm), creep strain rate (mm/mm/hrs), and rupture elongation (%) of type 316SS are higher than that of type 316LN SS for the same temperatures and applied constant stress level. It is concluded that type 316LN SS has a good creep resistance than that of type 316SS.

LDA measurements of coherent flow structures and cross-flow across the gap of a compound channel with two half-rods

The enhancement of heat transfer from fuel rods to coolant of a Liquid Metal Fast Reactor (LMFR) decreases the fuel temperature and, thus, improves the safety margin of the reactor. One of the mechanisms that increases heat transfer consists of large coherent structures that can occur across the gap between adjacent rods. This work investigates the flow between two curved surfaces, representing the gap between two adjacent fuel rods. The aim is to investigate the presence of the aforementioned structures and to provide, as partners in the EU SESAME project, an experimental benchmark for numerical validation to reproduce the thermal hydraulics of Gen-IV LMFRs. The work investigates also the applicability of Fluorinated Ethylene Propylene (FEP) as Refractive Index Matching (RIM) material for optical measurements.The experiments are conducted on two half-rods of 15 mm diameter opposing each other inside a Perspex box with Laser Doppler Anemometry (LDA). Different channel Reynolds numbers between Re = 600 and Re = 30,000 are considered for each P/D (pitch-to-diameter ratio).For high Re, the stream wise velocity root mean square vrms between the two half rods is higher near the walls, similar to common channel flow. As Re decreases, however, an additional central peak in vrms appears at the gap centre, away from the walls. The peak becomes clearer at lower P/D ratios and it also occurs at higher flow rates. Periodical behaviour of the span wise velocity across the gap is revealed by the frequency spectrum and the frequency varies with P/D and decreases with Re. The study of the stream wise velocity component reveals that the structures become longer with decreasing Re. As Re increases, these structures are carried along the flow closer to the gap centre, whereas at low flow rates they are spread over a wider region. This becomes even clearer with smaller gaps.

Pressure Measurements in a Wire-Wrapped 61-Pin Hexagonal Fuel Bundle

To achieve longer-life liquid-metal fast reactor cores, designers are considering to increase the wall gap of the wire-wrapped hexagonal fuel bundles to account for volumetric void swelling and radiation creep. A new wire-wrapped hexagonal test bundle has been constructed, with a wall gap larger than prior experiments, and experimental pressure drop data have been generated under laminar, transition, and turbulent flow regimes (corresponding to Re of 250–19,000), to complement the existing database of small wall gap experimental bundles. The comparison of the experimental data set with the predictions of four existing correlations (Baxi and Dalle Donne, Cheng and Todreas detailed (CTD), Kirillov, and Rehme) showed general agreement between data and the selected correlations. However, the CTD correlation most accurately predicted the experimental trend and the transition between flow regimes. The analysis of the experimental data also revealed that the larger wall gap size caused a lower bundle pressure drop due to the increased bypass flow area.

A negative reactivity feedback driven by induced buoyancy after a temperature transient in lead-cooled fast reactors

Abstract Consideration is given to the possibility to use changes in buoyancy as a negative reactivity feedback mechanism during temperature transients in heavy liquid metal fast reactors. It is shown that by the proper use of heavy pellets in the fuel elements, fuel rods could be endowed with a passive self-ejection mechanism and then with a negative feedback. A first estimate of the feasibility of the mechanism is calculated by using a simplified geometry and model. If in addition, a neutron poison pellet is introduced at the bottom of the fuel, then when the fuel element is displaced upward by buoyancy force, the reactivity will be reduced not only by disassembly of the core but also by introducing the neutron poison from the bottom. The use of induced buoyancy opens up the possibility of introducing greater amounts of actinides into the core, as well as providing a palliative solution to the problem of positive coolant temperature reactivity coefficients that could be featured by the heavy liquid metal fast reactors.

On the use of a dedicated ballast pellet for a prompt self-ejection mechanism after a temperature transient in lead-cooled fast reactors

The potential use of changes in buoyancy as a reactivity feedback mechanism during temperature transients in heavy liquid metal fast reactors (HLMFRs) is discussed. It is shown that with the use of ballast pellets (∼15% volume fraction) introduced in combination with fuel pellets, fuel rods will be endowed with a reliable self-ejection mechanism that is able to compensate temperature transients. Utilizing a simplified model, an estimate of the negative reactivity insertion expected from this mechanism is derived. The use of ballast pellets opens up the possibility of introducing greater amounts of actinides into the core, as well as providing a solution to the classical problem of positive coolant temperature reactivity coefficients in fast reactors.

Investigation of molten fuel coolant interaction phenomena using real time X-ray imaging of simulated woods metal-water system

Abstract In liquid metal fast breeder reactors, postulated failures of the plant protection system may lead to serious unprotected accidental consequences. Unprotected transients are generically categorized as transient overpower accidents and transient under cooling accidents. In both cases, core meltdown may occur and this can lead to a molten fuel coolant interaction (MFCI). The understanding of MFCI phenomena is essential for study of debris coolability and characteristics during post-accident heat removal. Sodium is used as coolant in liquid metal fast breeder reactors. Viewing inside sodium at elevated temperature is impossible because of its opaqueness. In the present study, a methodology to depict MFCI phenomena using a flat panel detector based imaging system (i.e., real time radiography) is brought out using a woods metal-water experimental facility which simulates the UO 2 -Na interaction. The developed imaging system can capture attributes of the MFCI process like jet breakup length, jet front velocity, fragmented particle size, and a profile of the debris bed using digital image processing methods like image filtering, segmentation, and edge detection. This paper describes the MFCI process and developed imaging methodology to capture MFCI attributes which are directly related to the safe aspects of a sodium fast reactor.

Risk assessment of thermal striping in MYRRHA research reactor

The thermal striping is one of the problems encountered in nuclear facilities. It may affect the reactor main components structural integrity and, sometimes, it can cause their rapid failure.The thermal striping problem touches, in general, mixing zones of industrial facilities. While it affects several areas as the core of liquid metal reactors that is due to the larger difference between core inlet and core outlet temperature, these reactors are more vulnerable to striping compared to water reactors.It is true that a better understanding of the induced thermal loads and associated damage mechanisms are obtained nowadays, thanks to recent developments in different domains such as CFD calculation and fatigue modelling. However, in the MYRRHA reactor project, the coolant used will be the Lead Bismuth Eutectic (LBE), i.e. a liquid-metal with a high thermal conductivity. Such conditions are broadly different compared to the most conventional coolant that is water.The present paper presents an assessment of the consequences of thermal striping under the severe conditions of the MYRRHA reactor. The results of the study can be considered as satisfactory since the risk to structural integrity appears to be globally low. A number of aspects will however require further investigation as the reactor design progresses.

Analyses of deformation and thermal-hydraulics within a wire-wrapped fuel subassembly in a liquid metal fast reactor by the coupled code system

The coupled numerical analysis of mechanical and thermal-hydraulic behaviors was performed for a wire-wrapped fuel pin bundle subassembly irradiated in a fast reactor. For the analysis, the fuel pin bundle deformation analysis code BAMBOO and the thermal-hydraulic analysis code ASFRE exchanged the deformation and temperature analysis results through the iterative calculations to attain convergence corresponding to the static balance between deformation and temperature. The analysis by the coupled code system showed that the radial distribution of coolant temperature in the subassembly tended to flatten as a result of the fuel pin bundle deformation governed by cladding void swelling and irradiation creep. Such flattening of temperature distribution was slightly observed as a result of fuel pin bowings due to the cladding-wire interaction even when no bundle-duct interaction occurred. The effect of the spacer wire-pitch on deformation and thermal-hydraulics was also investigated in this study.

Numerical simulations on self-leveling behaviors with cylindrical debris bed

The postulated core disruptive accidents (CDAs) are regarded as particular difficulties in the safety analysis of liquid-metal fast reactors (LMFRs). In the CDAs, core debris may settle on the core-support structure and form conic bed mounds. Then debris bed can be levelled by the heat convection and vaporization of surrounding coolant sodium, which is named “self-leveling behavior”. The self-leveling behavior is a crucial issue in the safety analysis, due to its significant effect on the relocation of molten core and heat-removal capability of the debris bed. Considering its complicate multiphase mechanism, a comprehensive computational tool is needed to reasonably simulate transient particle behavior as well as thermal-hydraulic phenomenon of surrounding fluid phases. The SIMMER program is a successful computer code initially developed as an advanced tool for CDA analysis of LMFRs. It is a multi-velocity-field, multiphase, multicomponent, Eulerian, fluid dynamics code coupled with a fuel-pin model and a space- and energy-dependent neutron kinetics model. Until now, the code has been successfully applied in numerical simulations for reproducing key thermal-hydraulic phenomena involved in CDAs as well as performing reactor safety assessment. However, strong interactions between massive solid particles as well as particle characteristics in multiphase flows were not taken into consideration in its fluid-dynamics models. To solve this problem, a new method is developed by combining the discrete element method (DEM) with the multi-fluid models of SIMMER code to reasonably simulate the particle behaviors as well as the thermal-hydraulic phenomena of multiphase fluid flows. In this paper, the coupling method is validated by performing numerical simulations on a series of experiments with cylindrical particle bed in 3D situation. Based on various experimental conditions, Reasonable agreement between simulation results and corresponding experimental data can demonstrate the applicability of the method in reproducing the self-leveling behavior of debris bed. Sensitivity analysis on some model parameters of DEM is also performed to assess their impacts in the simulation. It is expected that the present method can be used as a computational tool to estimate self-leveling process of debris beds as well as the following behaviors in real reactor environment.

Computational simulation of thermal hydraulic processes in the core and fuel assembly in the liquid-metal fast-breader reactor in the porous medium approximation

The results of numerical simulation of the sodium coolant flow in the fuel assembly with partial blocking of its cross section are presented. A comparison of the results obtained with experimental data confirmed the operability of the APMod module intended for simulating heat exchange processes in cores and heat exchange equipment of promising nuclear reactors.