Sodium-cooled Fast Reactor Research Articles

How could fuel corrosion influence the delayed neutron signal time evolution in sodium cooled fast reactors?

Sodium cooled fast reactors (SFR) employ a fuel pin failure detection system to limit the release of fissile material into the coolant. The system monitors for a neutron signal in the coolant resulting from fission products released by cladding failure. The time evolution of the neutron signal in a delayed neutron detection system (DND) is related to the release mechanisms of delayed neutron precursors (DNP) as well as the fuel pin evolution. In order to interpret the delayed neutron signal’s time evolution and better identify the breached fuel area, the enhanced thermal diffusion release due to corrosion layer formation during the fuel evolution is modeled in this paper. As an experimental benchmark, one experimental case (RG 11 in the PHENIX reactor) is studied in this paper. By modeling the diffusive release fraction during corrosion layer formation and comparison with experimental data, we deduce that the enhanced diffusive release, which is influenced by temperature changes at the fuel rod periphery due to the corrosion layer formed, provides a good interpretation for the time evolution of the delayed neutron signal. In addition, the temperature rise due to the corrosion layer may also cause a pulsed gas release and a corresponding delayed neutron signal increase.

3D simulation of transient thermal-hydraulic characteristics in CEFR during emergency shutdown event

Safe shutdown behavior of China Experimental Fast Reactor (CEFR) in emergency status is one of the key points in its safety assessment. The CEFR design incorporates a pool-type structure. However, the flow and heat transfer characteristics in the pool are complex, which results in obvious three-dimensional (3D) effects. Therefore, obtaining the 3D flow paths and accurate temperature distribution characteristics using a one-dimensional system code is difficult. In the present work, 3D computational fluid dynamics simulations were performed on the pool-type sodium-cooled fast reactor CEFR in full-scale. The “decoupling modeling and coupling calculation” method was used for performing numerical simulation to comprehensively investigate the 3D thermal and flow characteristics under emergency shutdown conditions. The convection and the corresponding heat transfer effects were evaluated by investigating both inter-flow and intra-flow paths in the reactor core. FLUENT 18.0 results were compared and verified with system code SELTAC simulation results. The results revealed that the residual heat was mainly dissipated by the intra-subassembly flow under reactor emergency shutdown conditions. The maximum core outlet temperature did not exceed 543.4 °C under emergency shutdown event, which is lower than the safety limit. This phenomenon indicates that if the heat-removal capability is maintained in the main heat exchange system, CEFR shutdown protection is effective. The three-dimensional results of the present work provide important numerical references for CEFR safety analysis.

Qualification of Metallic Fuel Data for Advanced SFR Applications

Metallic fuel is proposed for use in various advanced sodium-cooled fast reactor (SFR) designs, including small modular reactors and microreactors. To support advanced SFR research and development, a metallic fuel database has been developed at Argonne National Laboratory. The database was built on extensive metallic fuel data collected during the Integral Fast Reactor Program, mainly from irradiation experiments in the Experimental Breeder Reactor-II (EBR-II). In order to manage the large amount of data and associated documents and to qualify the data, a Quality Assurance Program Plan (QAPP) was established for the legacy metallic fuel data following the American Society of Mechanical Engineers (ASME) Nuclear Quality Assurance (NQA-1) 2008/2009 Addenda Standard. Metallic fuel data qualification is being performed following the QAPP. The plan has been approved by the U.S. Nuclear Regulatory Commission and can be referenced in licensing applications for nuclear power plants. The process for qualifying metallic fuel data is presented. The structure for implementing the QAPP and the qualification methods are introduced. The principle for prioritizing the metallic fuel data for quality assurance is summarized. An overview of the metallic fuel data available and considered for qualification is also provided.

Nickel-free austenitic stainless steel alloys as a sodium-cooled fast reactor fuel cladding

ABSTRACT Nickel-free austenitic stainless steel alloys Fe-Cr-Mn (SSMn) and Fe-Cr-Mn-N (SSMnN) were successfully developed to be used as sodium-cooled fast reactor fuel cladding. An austenite phase was observed through microstructure examination for both SSMn and SSMnN. The effect of Mn and Mn-N as an alternative to Ni on tensile properties was studied at room temperature. Subsequently, the effect of temperature was studied through tensile testing of the alloys produced at 300 ºC and 500 ºC. Mn and Mn-N enhance the room temperature yield strength (YS) and ultimate tensile strength (UTS). Mn caused a small reduction in elongation (El%) but Mn-N caused its deterioration. The YS, UTS, and El% of both SSMn and SSMnN alloys are strongly temperature dependent. Both the SSMn and SSMnN alloys showed better mechanical properties than the standard AISI304 at elevated temperatures. So, the SSMn and SSMnN alloys could be considered strong candidates as fuel cladding for sodium-cooled fast reactors.

Comprehensive analysis of the cladding tubes manufactured by a new 10Cr1SiY ferrite/martensitic steel

To improve the corrosion resistance of cladding tubes used in advanced nuclear reactors such as lead fast reactor (LFR), sodium-cooled fast reactor (SFR), and super-critical water-cooled reactor (SCWR), 10Cr1SiY ferrite/martensitic (F/M) cladding tubes with 6.00 mm outer diameter, 0.5 mm wall thickness and 285–297 cm length were fabricated by cold-rolling, annealing, and designed normalizing and tempering (N&T) treatment before the last cold-rolling. The results showed a precise dimension along the tubes with an ovality lower than 1%. After 18.6% cold-rolling, isotropic structure with less differential of microstructure between the rolling section and cross section was observed by electron backscattered diffraction (EBSD) analysis. The slight rotation of martensite laths and formation of dislocation cells near ferrite borders would take responsibility for the phenomenon based on the coincidence site lattice (CSL) boundaries and orientation distribution function (ODF) plots. Ultimate tensile strengths (UTS) are 853 and 200 MPa while elongations are 7.8% and 27.5% for the 10Cr1SiY tube at room temperature (RT) and 700℃, respectively. The relationship between tube fracture surface and necking was found. A higher fraction of inner segment means more serious necking during the tensile test, which explains the higher total elongation but lower uniform elongation at 700℃.

Numerical study of gas-injection induced pool sloshing behavior using MPS method

When a Core Disruptive Accident (CDA) occurs in a Sodium-cooled Fast Reactor (SFR), as the accident progresses, its core may form a pool of molten fuel. When the molten fuel pool expands, a certain amount of coolant may be entrained in it. Because of the Fuel-Coolant Interaction (FCI), the molten fuel pool will have centralized sloshing behavior, making the fuel distribution more dense, leading to the risk of re-criticality, therefore, the study of centralized sloshing behavior is of great significance for evaluating the impact of the CDA. Under various experimental conditions, the sloshing experiment of gas injection can effectively simulate the mechanism and characteristics of centralized sloshing behavior. In this study, numerical simulation of gas-injection induced sloshing behavior is performed based on the Moving Particle Semi-implicit (MPS) method. Simulation results are compared with the experimental results. The results show that the deformation of bubble shape and liquid level in the sloshing characteristics are in good agreement with the experimental results. Therefore, the numerical simulation method established in this study can be used to study the pool sloshing characteristics induced by gas injection.

A study of the behavior of the floating absorber for safety at transient (FAST) in an innovative sodium-cooled fast reactor (iSFR)

Negative reactivity feedback is one of the most important inherent safety characteristics of reactors for suppressing changes in thermal power and component temperatures during accidents. The floating absorber for safety at transient (FAST) has been suggested to improve the coolant temperature coefficient (CTC) in a sodium-cooled fast reactor (SFR) for a better self-stabilization of the reactor system. Previous studies on the FAST application showed that the FAST causes power excursion and temperature oscillations under anticipated transient without scram (ATWS) conditions. This study explains the variables affecting the FAST behavior and oscillations with mathematical expressions. In addition, a damping FAST is proposed to passively remove oscillations without any mechanical system. The damping FAST increases the damping ratio by reducing the pin diameter at a height where the difference between the FAST bottom elevation and the equilibrium position is maximum. The performance evaluation for the damping FAST is conducted in an innovative sodium-cooled fast reactor (iSFR) under ATWS conditions. The 7-mm damping FAST unilaterally inserts the negative reactivity in unprotected transient overpower event, dramatically lowering the core power without oscillations. While the previous FAST shows oscillatory behavior repeatedly approaching the coolant boiling temperature, the damping FAST stably keeps the coolant temperature below 800 °C. Moreover, the damping FAST lowers the maximum coolant temperatures for other ATWS events. These findings show the potential of the FAST system to significantly improve the inherent safety of SFRs.

Experimental and numerical investigations into molten-pool sloshing motion for severe accident analysis of sodium-cooled fast reactors: A review

Safety issues are particularly crucial for sodium-cooled fast reactors (SFRs). Data obtained from SFR safety analyses over recent years have shown that a specific type of sloshing motion probably occurs in the molten pool during core disruptive accidents (CDAs) of SFRs due to local neutronic power excursion or pressure developments, thereby significantly influencing recriticality. Recognizing the importance of improving the evaluation of CDAs of SFRs, extensive knowledge about this phenomenon has been garnered through experimental studies of their thermal-hydraulic mechanism and characteristics. Based on these studies, simulations using various numerical approaches, such as SIMMER code, the finite volume particle method, and the smoothed particle hydrodynamic method, have attempted to reproduce the sloshing motion under various experimental conditions to verify their reasonability and applicability, thereby promoting the development of SFR safety analysis. To provide useful references for future SFR safety analyses and assessments, we have systematically reviewed and summarized these experimental and numerical investigations into the thermal-hydraulic aspect of molten-pool sloshing motion. In addition, to enhance deeper and more comprehensive research into sloshing motion, we have also discussed future prospects. Knowledge gained from experimental and numerical investigations into molten-pool sloshing motion is valuable not only for improving and verifying SFR safety analysis codes but also for providing reference data for studies of sloshing motion in other fields of engineering.

Debris Bed Self-Leveling Mechanism and Characteristics for Core Disruptive Accident of Sodium-Cooled Fast Reactor: Review of Experimental and Modeling Investigations

Evaluations of the Core Disruptive Accident (CDA) are significantly important for safety analysis of Sodium-cooled Fast Reactor (SFR) despite the very low probability of occurrence for CDA. During the material-relocation phase in CDA of SFR, the molten materials are possibly released from the core region into subcooled sodium, subsequently forming the debris bed on the lower part of the reactor vessel after being quenched and fragmented. The accumulated high-temperature debris with decay heat can cause sodium coolant boiling, leading to the so-called “debris bed self-leveling behavior” during which the shape of the debris bed becomes flattered (leveling). It is important to investigate the debris bed self-leveling behavior due to its potential capacity to induce the transfer of debris and affect the ability of cooling and criticality of the debris bed. Thus, in recent years, valuable knowledge concerning the mechanism and characteristics of this behavior was accumulated through lots of experimental results and modeling developments. Aimed at providing a valuable guideline for future investigations on this issue, in this study, the past experimental and modeling investigations on debris bed self-leveling mechanism and characteristics are systematically summarized and reviewed, and some future remarks are also proposed to promote the progression of further research for SFR severe accident analysis.

A 3D particle-based simulation of heat and mass transfer behavior in the EAGLE ID1 in-pile test

The ID1 test was the final target test of the EAGLE experimental framework program. It was used to verify that during a core disruptive accident, the molten fuel could be discharged via wall failure of an inner duct in FAIDUS, a design concept for the sodium-cooled fast reactor. The ID1 results revealed that the wall failure behavior owed to the large heat flow from the surrounding fuel/steel mixture. The present study numerically investigated the heat transfer mechanisms in the test using the finite volume particle method in the three-dimensional domain. The thermal hydraulic behaviors during wall failure were reproduced reasonably. The present three-dimensional simulation mitigated inherent defects of our previous two-dimensional calculation and clarified that the solid fuel and liquid steel close to the outer surface of the duct can expose the duct to high thermal loads, resulting in the wall failure.

Numerical study of gas bubble rising in liquid sodium using advanced MPS method

In Sodium cooled Fast Reactor (SFR), study on the behavior of gas bubble is very crucial to the safety of SFR. In this study, an advanced Moving Particle Semi-implicit (MPS) method, the least square MPS (LSMPS), is used to study the rising behavior of gas bubble in liquid sodium. To improve the calculation, modified multiphase model is developed for the purpose of clear interface capture. The accuracy of LSMPS with modified multiphase model in simulate single rising bubble behavior is verified by several benchmarks with available experiment results. Then LSMPS is used to study single gas bubble rising in liquid sodium pool with different diameters. Velocity and shape of gas bubble is predicted using the LSMPS method. Side wall effects is also investigated.

Effect of Internal Decay Power and Surface Cooling on Nuclear Fuel Droplet Solidification During MFCI in a SFR—A Numerical Study

Melting of the nuclear core is one of the severe accident scenarios in a Sodium-cooled Fast Reactor (SFR). During such an event, molten corium may come into contact with the coolant sodium. This interaction of the molten fuel and the coolant is commonly termed molten fuel–coolant interaction (MFCI) in the nuclear industry. In this study, a numerical analysis is carried out to study the solidification of a molten fuel droplet in the liquid sodium pool. In the first part of the study, the effect of constant internal heat generation on the solidification of the droplet is evaluated with convective heat dissipation prescribed at the droplet surface. The internal heat generation (decay power) and the heat transfer coefficient are varied as parameters, and the time required for complete solidification of the molten droplet is obtained. Based on the results, the freezing of the droplet is categorized into three regimes: conduction limited, transition, and internal heat generation dominated regimes. It is observed that the solidification process of nuclear fuel droplets generated during MFCI is not influenced by internal heat generation and lies in a conduction-limited regime for decay power level prevailing in a medium-sized SFR. Hence, in the next part of the study, the numerical analysis is carried out by incorporating the time-dependent decay power and the temperature-dependent heat transfer coefficient in the computational model by developing user-defined subroutines depicting a realistic scenario of an accident. The results of the analysis show that because of the high subcooling of sodium, film boiling is ruled out; nucleate boiling with a maximum heat transfer rate occurs briefly. The heat transfer coefficient then declines as the interface temperature between the droplet and the sodium decreases rapidly until the natural convective regime is reached. A parametric study on the droplet diameter is also carried out by varying the diameter from 0.5 to 10 mm, spanning the typical particle size spectrum expected during MFCI.

Versatile Test Reactor Conceptual Core Design

The VTR is a 300-MW(thermal) sodium-cooled fast reactor (SFR) designed for the specific purpose of delivering unique testing capabilities to enable the advancement of all reactor technologies. With its flux level, irradiation volume, and operational flexibility, the VTR will enable accelerated testing of materials, fuels, and various components needing irradiation testing. Proven SFR technologies and design approaches have been leveraged in designing the VTR core, ensuring the highest possible readiness level. This resulted in the VTR using ternary metallic fuel and delivering fast flux levels in excess of 4 × 1015 n/cm2·s over large useful volumes, corresponding to about 60 dpa/year in steel. As part of the design efforts, the VTR core performance has been determined for a representative configuration, ensuring that the reactivity control systems offer sufficient shutdown margins, that the core can be safely cooled in all situations, and that reactivity feedback coefficients are conducive to a favorable safety behavior. Furthermore, the incorporation of features such as fuel assembly storage in the shield region supports the flexible and reliable operation of the VTR. Additional design work has been ongoing as well. This includes thorough shielding performance evaluations to ensure safe operation of the VTR, verification and validation of the design tools used to achieve compliance with Nuclear Quality Assurance (NQA-1) requirements, early assessment of the impact of irradiation experiments on the core performance envelope and associated margins, and in-depth uncertainty quantification efforts to quantify the anticipated range of performance characteristics. An experimental program supporting the VTR core design has been set up, with the current focus being on thermal-hydraulic experiments. The purpose of this experimental program is to obtain confirmatory measurements to serve directly as part of the core design basis or as part of the validation cases supporting the simulation tools used.

Performance test and uncertainty analysis of the FBG-based pressure transmitter for liquid metal system

The pressure measurement in the high-temperature liquid metal system, such as Sodium-cooled Fast Reactor(SFR), is important and yet it is very challenging due to its nature. The measuring pressure is relatively at low range and the applied temperature varies in wide range. Moreover, the pressure transfer material in impulse line needs to considered the high temperature condition. The conventional diaphragm-based approach cannot be used for it is impossible to remove the effect of thermal expansion. In this paper, the Fiber Bragg Grating(FBG) sensor-based pressure measuring concept is suggested that it is free of problems induced by the thermal expansion. To verify this concept, a prototype was fabricated and tested in an appropriate conditions. The uncertainty analysis result of the experiment is also included. The final result of this study clearly showed that the FBG-based pressure transmitter system is applicable to the extreme environment, such as SFR and any other high-temperature liquid metal system and the measurement uncertainty is within reasonable range.

Transmission electron microscopy study of a high burnup U-10Zr metallic fuel

To support the development of U-10 wt.% Zr (U-10Zr) metallic fuel for Generation IV sodium-cooled fast reactors, we analyzed a Na-bonded solid U-10Zr fuel cross-section that was irradiated to a burnup of approximately 13.2 at.% at the Fast Flux Test Facility (FFTF). Advanced characterization techniques, including site-specific sample preparation by focused ion beam (FIB) and scanning transmission electron microscopy (STEM), were used to reveal the Zr redistribution and characterize the fuel matrix and secondary phases (such as solid fission products) present at the end of life. Results showed that the fuel pin cross-section is divided into three major concentric zones: a Zr-rich central region, a Zr-lean intermediate region, and a Zr intermediate peripherical region. The phase characterization revealed that the irradiation environment enhanced the development and stabilization of phases not predicted by the standard equilibrium U-Zr phase diagrams. Comparing the current results with the ones from previous studies, it is reaffirmed that the radial temperature profile and the time spent in the reactor, rather than the fuel burnup, are the two factors that most influence the formation of redistribution zones and their extension along the fuel cross-section. Various solid fission products, such as lanthanides, ZrRu, BaTe, CsI, and Ba and Sr oxides, precipitated inside the fission gas pores. This study provides unprecedented nanoscale understandings in the irradiated U-10Zr fuel system that may benefit fuel performance modelling and advanced fuel development.

Metal fuel relocation experiments with pressure injection

A pool-type sodium-cooled fast reactor (SFR) using metal fuels has a number of inherent safety features that can support benign consequences for design basis accidents (DBAs). Even in postulated severe accident conditions, the core is designed to remain under sub-critical condition in a passive coolable geometry. In case of the postulated severe accidents in SFRs, fuel relocation in the core region along the coolant channel is an important negative reactivity feedback factor that lowers the reactor power level and consequently eliminates the possibility of recriticality. Therefore, understanding the relocation behavior of fuels and the coolability of the relocated fuels in the postulated severe accident is one of the most important factors in the safety assessment of SFRs. In the present study, the relocation behavior of the metal fuel in a pin bundle geometry was investigated by injecting the metal fuels into the coolant channels with pressure. The first metal fuel relocation with pressure injection (RPI-1) experiment showed that many of the metal fuels levitated to upper plenum. In case of RPI-2 experiment, the Real Time X-ray Video System (RTXVS) was constructed and used to obtain a real time X-ray video of the experiment. This real time X-ray video clearly showed the relocation behavior of the pressure injected metallic uranium in the sodium coolant channel. Through this video, it was confirmed that part of the fuel was dispersed downward and part of the fuel was dispersed upward, and the relocation behavior of the injected fuel proceeded simultaneously in the downward and upward directions. It can be concluded that in case of pressure injection of the metallic fuel, there is a possibility that the metallic fuel can be quickly removed from the core region, resulting in a negative reactivity feedback effect.

FAST irradiations and initial post irradiation examinations – Part I

The Advanced Fuels Campaign Fission Accelerated Steady-state Test (FAST) at Idaho National Laboratory (INL) completed its first irradiation cycle within the Advanced Test Reactor (ATR). The test focused on the irradiation of alloy fuel forms for use in sodium fast reactors. The first cycle of FAST testing was completed and four rodlets were removed for the initial post irradiation examination (PIE). The rodlet design and irradiation conditions were evaluated using Monte Carlo N-Particle (MCNP) for as-run power history and COMSOL for temperature analysis. These rodlets include a set of low burnups (∼2.5 % fissions per initial metal atoms [%FIMA]), control rodlets, and a helium-bonded annular rodlet (4.7 %FIMA). Non-destructive PIE has been completed and includes visual inspection, neutron radiography and gamma scanning of the FAST capsules and rodlets. Radiography confirmed the integrity of the experiments, revealed that the annulus in the annular fuel was filled at a modest burnup (4.7 %FIMA), and indicated potential slumping of the cooler rodlets at lower burnup. Precision gamma scanning indicated mostly usual fission product behavior, except for cesium in the He-bonded annular fuel. Future destructive PIE will be necessary to fully interpret the effects of accelerated irradiation on U-Zr metallic fuel behavior.

Computational Aspects of Core Neutronics Calculation of Sodium-Cooled Fast Reactor at the Core Destruction Stage

Computational Aspects of Core Neutronics Calculation of Sodium-Cooled Fast Reactor at the Core Destruction Stage

Evaluation of Microstructural Evolution in Isothermally Aged Ferritic Candidate Cladding Materials for Sodium-cooled Fast Reactor Applications

Evaluation of Microstructural Evolution in Isothermally Aged Ferritic Candidate Cladding Materials for Sodium-cooled Fast Reactor Applications

Study on criteria and prediction method of liquid fall type gas entrainment in pool-type sodium-cooled fast reactors

Liquid fall type entrainment, a kind of gas entrainment phenomenon which may undesirably occur in sodium-cooled fast reactors (SFRs) due to complex immersed structures and high-velocity flowing coolants of SFRs, will involve bubbles into coolants and lead a series of safety risks. Aiming at providing a guideline for avoiding this phenomenon, the computational fluid dynamics (CFD) method was adopted to investigate the influences of fluid properties on entrainment initiation. Before the simulation case study, a small-scale experiment was conducted for verification which demonstrated the numerical model and VOF method are suitable for this study. The modeling criteria of simulating sodium entrainment by other liquids were determined by simulation results. It was also found that the relation between critical inlet velocity and σ/ρ, and the relation between critical re-submergence velocity and critical inlet velocity can be accurately estimated by the case validation. Based on which, an improved wide-applicable prediction method was proposed for entrainment situation judgment.