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• Steam generators (SG) in SFRs are operated at higher temperatures and higher pressure margins than the LWR counterparts. • Other failures, e.g. fatigue, stress corrosive crack, fretting, etc., are far less likely than creep during normal operation. • Creep dominates SFR SG degradationsgiven the high temperature and can be described by Larson-Miller parameter (LMP). • The proposed prognostic model estimates the failure probabilities by sampling LMR and temperature and their uncertainties using Monte-Carlo methods. • The model successfully predicts the failure probability within a given time and suggests how adjusting the workload can protect the component. This paper presents a prognostic model for sodium-cooled fast reactor (SFR) steam generators (SGs). The purpose of the model is to estimate the remaining useful life of SFR SGs and thus to support the decision-making of autonomous control. SFR SGs are of great interest for plant integrity due to their harsh operation environment. They operate at higher temperatures and higher coolant-to-steam pressure differences than those of current light water reactors (LWR). The severity of the SFR SG failure consequences, which include water-sodium contact, is another reason. Understanding its failure mechanisms is important for the development of its prognostic model. Based on our literature review and physics-based analysis, we concluded that creep would be a dominant degradation mode of SFR SGs due to SFR's high-temperature and high-pressure environment. Thus, creep is the focus of the prognostic model. Various other failure modes were also investigated in this study. The mechanical fatigue due to flow-induced vibration is usually observed in early developmental phases and would not be an important issue during normal operation. The thermal fatigue due to thermal stripping is occasionally observed in other components in SFRs but does not affect SG integrity. Pure water stress corrosive crack, fretting, and so on are commonly observed in LWR SGs but are significantly less important in SFR SGs because of the high temperatures, high pressure differences, and chemical properties of liquid sodium. Based on these investigations, a prognostic model focusing on creep failures was developed. It estimates the failure probability profile by sampling the Larson-Miller parameter (LMP) and temperature and associated uncertainties through Monte Carlo methods. Two case studies were presented. The first one demonstrated the model's capability to calculate the failure probability for a new specimen within a given time. If the tube is working under temperatures of 500C ± 3 and pressures of 15.2 MPa, which leads to LMP of 20,100 ± 50, the failure probabilities within 50, 70, and 90 years are approximately 0.1 %, 3.2 %, and 17.5 %. The second demonstrated how adjusting the workload can help to protect the integrity of the component. For an old tube reaching 99 % of its lifespan, continuing to run at temperatures of 500C ± 3 and LMP of 20,100 ± 50 leads to a failure probability of about 29.0 % within a year. If the temperature is reduced by 5C, the failure probability can be reduced to 2.5 %, and, if the pressure is also reduced such that the LMP is increased by 100, the failure probability can further be reduced to 0.12 %. [ABSTRACT FROM AUTHOR]

• Demonstration plant of sodium-cooled fast reactor. • A design and lifecycle simulation/evaluation system. • Safety design and evaluation methods. • Codes and standards taking into account transformation to risk-informed and performance-based framework. • Separation and transmutation of minor actinides in nuclear fuel cycle. A sodium cooled fast reactor (SFR) is one of the most relevant and decarbonized energy supply system with higher sustainability on natural resources, footprint, and waste management. It was planned in a "strategic roadmap" of fast reactor decided by Inter-Ministerial Council for Nuclear Power Japan in 2022 to start a conceptual design of a demonstration reactor from 2024 with a background of accumulated knowledge and experiences of SFR development. For example, a design and lifecycle simulation/evaluation system named ARKADIA has been developed to accelerate such design works. It will enable to take into account plant lifecycle, e.g., operation and maintenance, to the plant design and optimize it based on simulations and knowledgebase. This paper shows research progresses of ARKADIA, safety design and evaluations, codes and standards, fuel cycle, and SFR development projects in Japan. [ABSTRACT FROM AUTHOR]

• Past findings on debris bed behaviors are summarized for enhanced SFR safety. • Debris bed formation & self-leveling mechanisms and characteristics are clarified. • Study at more realistic conditions and future development direction are prospected. Understanding the Core Disruptive Accident (CDA) is crucial for the safety analyses of Sodium-cooled Fast Reactors (SFR), despite its exceedingly low possibility. During a CDA, the potential release and subsequent relocation of molten materials from the core could result in the formation of a debris bed inside the reactor vessel due to their quenching and fragmentation from the interactions with the subcooled sodium coolant. Debris beds with varying shapes are expected to form due to the different flow-regime characteristics during their formation. Further, the heat decay from this high-temperature debris might trigger sodium coolant boiling, instigating the debris bed self-leveling behavior, which flattens the debris beds. Considering the importance of the debris bed shapes to the debris transfer, heat removal and neutronic criticality, lots of knowledge about the mechanisms and characteristics of debris bed formation and self-leveling behaviors has been gathered in recent years through extensive experimental, modeling and numerical outcomes. This study aims to systematically and critically summarize these past explorations into the debris bed formation and self-leveling behaviors, offering a beneficial direction for future research on the improved safety evaluations in SFR severe accidents. [ABSTRACT FROM AUTHOR]

[Display omitted] • GRESS code is developed for analyzing fission gas release and swelling in SFR. • Grain size and gas bubble diameter exhibit rapid variations within a specific range of fuel temperature. • The relatively minimal variation in grain size and gas bubble diameter at higher fuel temperatures result in a decrease in gas bubble swelling. • FGR and intragranular gas bubble swelling increase over time during transient operation. • The results demonstrate the coupling and interaction of fission gas behavior with each other. In order to simulate the behaviors of Fission Gas Release (FGR) and swelling for Sodium-cooled Fast Reactor (SFR) under normal and transient conditions, a G as RE lease and S welling S imulations (GRESS) code is developed using COMSOL Multiphysics. GRESS is mainly based on a previous model called the Operational Gas RElease and Swelling (OGRES) model, in which the FGR and swelling are primarily related to burnup, time and temperature. GRESS assumed that there is a spherical grain at the fuel pellet. The grain is presented by a 2D axisymmetric geometry in COMSOL Multiphysics. Fission gas atom generation, gas atom diffusion, gas atom diffusion into the bubble, bubble resolution, gas bubble nucleation, gas bubble migration to the grain boundary, bubble pressure, gas bubble swelling, grain growth, oxygen to metal atomic ratio, fuel porosity, fuel thermal conductivity, heat generation rate, pore velocity, pore concentration, plutonium migration and so on are considered in GRESS code. The output results include grain size, gas bubble diameter, gas bubble concentration, the amount of gas atoms per bubble, gas atom diffusion coefficient, bubble diffusion coefficient, surface diffusion coefficient (varying with temperature), FGR (varying with time), and swelling (varying with time and radius). The comparison of the results with other modeling outcomes and experiments is performed, which shows a good agreement. [ABSTRACT FROM AUTHOR]

• The thermal stress induced cracking of nuclear fuel droplet fragmenting in sodium is studied. • Fracture Bifurcation Theory (FBT) and Minimum Potential Energy Principle (MPEP) approaches are adopted for the analysis. • Parametric study carried out for different droplet sizes, crust thickness and cooling conditions. • The minimum size of the droplet prone to catastrophic failure is found out using FBT. • Fine fragment detached from the solidified crust due to thermal stress cracking evaluated using MPEP. During Molten Fuel Coolant Interaction (MFCI) occurring in a Sodium-cooled Fast Reactor (SFR) following a severe accident, the corium jet breaks up into coarse droplets due to hydrodynamic instabilities and fine flaked debris are generated due to the thermal stress cracking occurring on the solidifying droplet. In particular mixed oxide core debris quenched by liquid sodium are more prone to thermal stress cracking because of the excellent heat transport properties of liquid sodium, which leads to a large temperature gradient across the crust of the ceramic fuel. In this paper, thermal stress initiated cracking of solid crust of the nuclear fuel droplet is studied using Fracture Bifurcation Theory (FBT) and Minimum Potential Energy Principle (MPEP). FBT is able to predict the minimum crack length needed for crack propagation or catastrophic failure depending on the computed K I values. On the other hand, MPEP can predict the periodic and hierarchical cracking pattern that can develop on the core debris. This helps in the prediction of number of cracks and the crack length that can develop on the solid crust. Parametric studies have been carried out by varying MOX droplet sizes, crust thickness and cooling conditions and calculating K I by semi-analytical means and comparing it with K IC. The analysis revealed that solidifying droplets with radius equal to or more than 2 mm are susceptible to thermal stress initiated crack growth and catastrophic breakage even if a few tens of micron size initial crack develops on its surface. Then a particular case of 2 mm radius core debris is considered and its cracking pattern is studied by applying MPEP. This optimal crack configuration is obtained by minimizing potential energy for a given range of crack length and spacing. The analysis has been carried out with ANSYS APDL code which calculates the strain energy for a given initial crack length and spacing. The 2D computational model is validated with a similar study carried out for Alumina disc quenched in water and the results are in good agreement. The minimum fine fragment size predicted by the analysis for mixed oxide debris is a flake like debris with equivalent radius of 41 µm when approximated to be a sphere. The particle size correlates well with the fine debris sizes reported in literature of MFCI experiments. [ABSTRACT FROM AUTHOR]

• Numerical simulation of debris bed formation is performed using MPS-DEM method. • High-order MPS method is applied with DEM model for the accurate simulations. • Applicability and reliability of the method are validated by experimental results. • Liquid pool convection induced by falling particles is confirmed in simulations. • Fuel debris bed formation in sodium is simulated to study for CDA analysis of SFR. During a Core Disruptive Accident (CDA) in a Sodium-cooled Fast Reactor (SFR), debris beds may form in the lower plenum of the reactor vessel as a result of the accumulation of discharged debris, which is formed from the solidification and fragmentation of core molten materials during their release due to interactions with the sodium coolant. Since the configuration of debris beds plays a vital role in their long-term coolability and neutronic subcriticality for achieving in-vessel retention, it is essential to understand and assess the Debris Bed Formation Behavior (DBFB). In this study, an advanced Moving Particle Semi-implicit (MPS) method, called least square MPS (LSMPS), is utilized and coupled with the Discrete Element Method (DEM) for the numerical investigations of DBFB characteristics in a two-dimensional liquid pool. The applicability and reliability of the proposed LSMPS-DEM method are validated by comparing simulation results with experiments using simulant materials. The LSMPS-DEM method is then employed to study DBFB in sodium. It is observed that debris properties significantly impact the sedimentation and accumulation processes of DBFB, leading to variations in the characteristics of the formed debris beds. The present study is expected to be useful for further investigations of DBFB under more realistic conditions and for the design improvement of safety devices to mitigate the consequences of CDA in SFR. [ABSTRACT FROM AUTHOR]

• A design analysis code for sodium heated OTSG is developed based on the power mesh method. • The code is verified with various design data both of integral type and modular type OTSGs. • The influence of tube diameter on heat transfer capacity has been analyzed by the code. Sodium heated once-through steam generator (OTSG) is a critical component of Sodium cooled Fast Reactor (SFR). It is a counter current heat exchanger. It transfers heat from hot sodium to water. Therefore, the steam generator is a necessity for the operation of the plant. However, due to the phase change exists in water side, it becomes challenging to design a steam generator for SFR. After the successful construction of CEFR (China Experiment Fast Reactor), a power-up demonstration SFR is being developed in China. For the design and analysis of the steam generator of the demonstration SFR, a thermal hydraulic Design Code of Once-through Steam generator for Sodium cooled fast reactor (DeCOSS) has been developed based on the movable mesh method. The DeCOSS code has been benchmarked based on the design data of OTSGs reported. The tube lengths and temperature distributions calculated by DeCOSS showed general agreements with published data. The influences of tube diameters on heat transfer capacity have been analyzed. The results indicated that the heat transfer capacity is determined by the thermal resistance of tube wall under the condition of fixed heat transfer coefficient of water side. In addition, under the same flow area of water and sodium, the heat transfer capacity could be enhanced by decreasing the tube diameters with constant tube wall thickness. Hence, the code can be utilized to do the sensitivity analyses and optimum thermal design of OTSGs. Moreover, it can support the design work of once-through steam generator of SFR. [ABSTRACT FROM AUTHOR]

• Test facility with 1/29.6 volume scale was built for simulating the tube bundle regions. • Test facility with 1/5 length scale was constructed for simulating the exit flow channel. • Pressure drop for the IHX shell side was determined by two different test facilities. • Pressure drop for the IHX was precisely measured over a wide range of flow rates. The flow characteristics at the shell side of a prototype intermediate heat exchanger (p-IHX) in a prototype generation-IV sodium-cooled fast reactor (PGSFR) were investigated experimentally in this study. The shell side of the p-IHX consisted of an inlet window, tube bundles, grid plates, and an exit flow channel. Two different test facilities were newly constructed to identify the pressure drop characteristics due to the complicated configuration with minimized scaling distortion. An intermediate heat exchanger test loop for PGSFR (iHELP) was constructed to characterize the flow resistance of the tube bundle regions with the grid plates and inlet window. The iHELP test section with a slab-shaped cross section was manufactured with a 1/29.6 vol scale ratio by preserving the height of the p-IHX. The hydraulic diameter at the tube bundle regions, the porosity of the grid plates, and the configuration of the tube bundles were conserved. In addition, the Reynolds number was preserved, which determined the flow rate conditions for the iHELP experiments. To investigate the pressure drop characteristics of the exit flow channel (EC), an intermediate heat exchanger test loop for the exit flow channel (IEC) was built with a 1/5 linearly reduced length scale by preserving the geometry of the flow path based on the EC of the p-IHX. The Euler number was conserved with a 1/8 Reynolds number ratio corresponding to a 1/1 flow velocity ratio to the p-IHX for dynamic similarity. By using each test facility the experimental pressure drop data satisfying the experimental uncertainty requirements were obtained separately for the bundle and exit channel regions. The pressure drop at the shell side of the p-IHX under a wide range of flow rate conditions was determined by using the pressure drop results from the iHELP and the IEC test facilities, which would be useful in improving and validating the pressure drop correlations in the IHX design computational code (SHXSA). [ABSTRACT FROM AUTHOR]

Highlights • A new method was proposed to simulate thermal deformation of subassembly in SFR. • Assembly deformation analysis code FADAC was developed and applied in the present study. • Free bowing was analyzed under different temperature conditions and displacement field of the subassembly was obtained. • Restrained bowing was analyzed in detail and maximum deflection was calculated. Abstract Due to the special geometry and compact arrangement of subassemblies in the sodium-cooled fast reactor, thermal deformation of the assemblies are easily triggered by uneven temperature distribution in the reactor core, which is negative for reactor safety. Therefore, it is necessary for researchers and engineers to conduct quantitative analysis towards thermal deformation behaviors of subassemblies, and furthermore, give reliable evaluation results on aftermath of the assembly thermal deformation. In the present study, thermal deformation analysis code FADAC, which was developed based on Euler-Bernoulli beam theory, was applied to make numerical investigation for thermal deformation behaviors of a single subassembly in sodium-cooled fast reactor. In order to preliminarily assess the capacity for thermal deformation, different temperature gradient conditions were considered and analyzed in detail for free bowing of the subassembly, furthermore, with the aid of numerical analysis for thermal axial forces and thermal bending moments, axial displacements and deflections were calculated finally. Besides, in order to predict the comprehensive results of the assembly deformation under both thermal load and mechanical load, restrained bowing were also analyzed for different temperature conditions. All of the simulation results were in good accordance with the experimental data. The present numerical research is of great significance to assembly deformation research in the sodium-cooled fast reactor, and will surely lay a solid foundation for deformation analysis towards multi-subassemblies. [ABSTRACT FROM AUTHOR]

Highlights • The deactivation process and nucleation process were analyzed. • Nucleation after interface movement can result in a larger value of γ. • A best-estimated correlation is proposed for calculation of nucleation radius. Abstract For the safety analysis of sodium-cooled fast reactor, nucleation superheat is a key parameter impacting the boiling process. For liquid metal sodium, models of conventional fluids do not apply and although much research has been performed, the basic mechanism has not been fully revealed. In this paper, by considering the effect of oxide layer and gas-liquid interface behavior in a conical cavity, the deactivation process and nucleation process were analyzed and a best-estimated correlation for nucleation radius was proposed. Taking the inert gas effect into account, the correlation was then used to predict the existing experiment data and good agreement was obtained. The results showed that there may be two types of nucleation due to the wetting property discrepancy, i.e. direct nucleation and the nucleation after interface movement, and the latter may be an explanation of the phenomenon of some data that the nucleation radius can be larger than the deactivation radius. It is also found that the ratio of deactivation radius and nucleation radius would vary with operation conditions and systems, and thus correlations should be employed rather than a constant value. [ABSTRACT FROM AUTHOR]

Highlights • Sloshing inside annular region separated by several cylinders was studied. • The frequency and wave height were relied on the number of internal cylinders. • The results using VOF approach were in good agreement with experimental results. Abstract The liquid sloshing behavior analysis of reactor vessel is an important topic of sodium-cooled fast reactor design. It is known that there are two pumps and four heat exchangers immersed in sodium liquid. Although lots of research work has been performed on the liquid sloshing inside simple cylindrical tank and coaxial circular cylinders, few about sloshing inside annular region separated by several cylinders (ARSSC), this paper deals with the analysis of the liquid sloshing inside the ARSSC geometry. A simplified cylindrical tank of fast reactor was designed, in which there are six internal cylinders used to model pumps and heat exchangers. By shaking table experiments, the number of internal cylinders was gradually increased, and the natural frequency and wave height of liquid sloshing were measured under each case. In numerical investigation, the VOF models were built to record wave height using the CFD simulation approach, the conditions were consistent with experiments. Much different with previous research, the characteristics of sloshing inside ARSSC geometry were strongly dependent on the number of internal cylinders. Experiment results showed that the natural frequency and wave height all decrease with the increase of internal cylinders gradually. The wave height response obtained by numerical simulation was in good agreement with the experimental results. In addition, as the increase of internal cylinders gradually, the maximum sloshing response occurred when the forcing frequency matched high order natural frequency of sloshing. Moreover, a correction factor was obtained by experiments for the natural frequency theoretical calculation formula. The conclusions can provide favorable reference for the seismic design of the sodium-cooled fast reactor. [ABSTRACT FROM AUTHOR]

Highlights • A detailed analysis of neutronics, core thermal-hydraulics, and fuel performance has been developed. • Thermo-mechanical failures of all the metallic fuel rods in a SFR are predicted. • This pin-level analysis leads to improved computational accuracy and provides more comprehensive physical information. Abstract A detailed analysis of neutronics, core thermal-hydraulics, and fuel performance has been developed to predict the thermo-mechanical failures of all the metallic fuel rods in a sodium-cooled fast reactor. The neutronics analysis provides a reload pattern for fuel assemblies and power/flux spatial distributions for each fuel rod. The core thermal-hydraulics allows the calculation of the cladding temperatures via subchannel analysis. The fuel performance analysis evaluates the thermo-mechanical integrity of all the fuel rods by considering the detailed spatial and temporal variations of the thermal power, neutron flux, and cladding temperature during the in-core lifetime of each fuel assembly. This analysis leads to improved computational accuracy and provides more comprehensive physical information for each code than a conventional one-dimensional fuel performance calculation. It is evident that, compared with a simple conservative one-dimensional analysis, the present method will improve plant performance while maintaining the thermal margin. The developed analysis has been applied to evaluate the fuel performance of a Prototype Gen IV sodium-cooled fast reactor candidate core by considering the uncertainties of the design parameters. [ABSTRACT FROM AUTHOR]

Two-phase flow through porous media must be well understood to develop a severe accident analysis code not only for light water reactor but also sodium-cooled fast reactor. When a core disruptive accident occurs in sodium-cooled fast reactor, the fuel inside the core become melted and interacts with the coolant. As a result, gas-liquid two-phase flow will be formed in the debris bed, which may have porous nature depending on the cooling process. In such condition, the local porosity and its distribution are very important to characterize two-phase flow field in the porous media. In this study, X-ray radiography was applied to measure the local porosity in the packed bed of spheres. The radial profiles were estimated from the chordal profiles measured by the X-ray method and compared with the previous porosity model. In addition, the void fraction radial profiles were also obtained in air-water two-phase flow. [ABSTRACT FROM AUTHOR]

• Discharge of sodium coolant into containment from a sodium-cooled fast reactor vessel can occur in the event of a pipe leak or break that can induce sodium fires. • MELCOR and SPHINCS, severe accident analyses codes model sodium fires in sodium-cooled fast reactor's containment. • To validate sodium pool fire models, experimental data (F7 series) from Japan Atomic Energy Agency was used. • Improvement on the current sodium pool fire model in MELCOR is also discussed. • Progress on the validation of MELCOR and SPHINCS is given. Discharge of sodium coolant into containment from a sodium-cooled fast reactor vessel can occur in the event of a pipe leak or break. In this situation, some of the liquid sodium droplets discharged from the coolant system will react with oxygen in the air before reaching the containment. This phase of the event is normally termed the sodium spray fire phase. Unreacted sodium droplets pool on the containment floor where continued reaction with containment atmospheric oxygen occurs. This phase of the event is normally termed the sodium pool fire phase. Both phases of these sodium-oxygen reactions (or fires) are important to model because of the heat addition and aerosol generation that occur. Any fission products trapped in the sodium coolant may also be released during this progression of events, which if released from containment could pose a health risk to workers and the public. The paper describes progress of an international collaborative research in the area of the sodium fire modeling in the sodium-cooled fast reactors between the United States and Japan under the framework of the Civil Nuclear Energy Research and Development Working Group. In this collaboration between Sandia National Laboratories and Japan Atomic Energy Agency, the validation basis for and modeling capabilities of sodium spray and pool fires in MELCOR of Sandia National Laboratories and SPHINCS of Japan Atomic Energy Agency are being enhanced. This study documents MELCOR and SPHINCS sodium pool fire model validation exercises against the JAEA's sodium pool fire experiments, F7-1 and F7-2. The proposed enhancement of the sodium pool fire models in MELCOR through addition of thermal hydraulic and sodium spreading models that enable a better representation of experimental results is also described. [ABSTRACT FROM AUTHOR]

To better understand the fundamental interactions between melt jet and coolant during a core-disruptive accident at a sodium-cooled fast reactor, the jet breakup and droplet formation in immiscible liquid-liquid systems were studied experimentally. Experiments using two different pairs of test fluids were carried out at isothermal conditions. The observed jet breakup behavior was classified into characteristic regimes based on the classical Ohnesorge classification in liquid-gas systems. The variation in breakup length obtained in the present liquid-liquid system was similar to that in a liquid-gas system. The droplet size distribution in each breakup regime was analyzed using image processing and droplet formation via pinch-off, satellite formation, and entrainment was observed. The measured droplet size was compared with those available from melt jet experiments. Based on the observation and analysis results, the breakup regimes were organized on a dimensionless operating diagram, with the derived correlations representing the criteria for regime boundaries of a liquid-liquid system. Finally, the experimental data were extrapolated to the expected conditions of a sodium-cooled fast reactor. From this, it was implied that most of the hydrodynamic conditions during an accident would be close to the atomization regime, in which entrainment is the dominant process for droplet formation. [ABSTRACT FROM AUTHOR]

• Lack of theory to accurately treat resonance self-shielding of displacement damage cross section. • Three approaches are proposed to compute resonance self-shielded damage cross section. • The self-shielding corrections on the displacement damage are close for the three methods. • The differences between methods are much smaller than the self-shielding corrections. • The main advantages and deficiencies of the three methods are discussed. Accurate calculation of neutron-induced displacement damage is a key step for the subsequent investigations on neutron irradiation damage. Due to the use of multigroup displacement damage cross sections, the resonance self-shielding effect of neutron-induced nuclear reaction must be taken into account. In addition to the method proposed by the authors in 2019, two additional methods are proposed in the present work to consider the self-shielding effect on the calculation of neutron irradiation-induced displacement damage. The two newly proposed methods allow directly using self-shielded multigroup nuclear reaction cross sections without infinite-dilution cross sections, even though they are not as consistent as the first one in the case of weak resonance self-shielding effect. The resulting resonance self-shielding corrections on the displacement damage of 56Fe in a typical sodium-cooled fast reactor are quite close for the three different methods. The slight differences (within ∼ 2 %) are much smaller than the self-shielding corrections, which are about −11.3 % for the dominant neutron elastic scattering-induced damage and −10.4 % for the total displacement damage. [ABSTRACT FROM AUTHOR]

• Unprotected Loss of Flow simulation of the European Sodium Fast Reactor. • Effect of sodium boiling has been analyzed on the reactor behavior with TRACE code. • Influence of a new subassembly design has been investigated on the boiling process. • Modified subassembly design shows enhanced safety characteristics during boiling. In this paper, an Unprotected Loss of Flow (ULOF) assessment has been performed on the European Sodium Fast Reactor developed in the ESFR-SMART EU project. To conduct the analysis, a simplified 42 channel thermal–hydraulic model has been established in TRACE system code, using a point kinetics model accounting for various reactivity feedback effects. The assessment reveals the core behavior of a commercial size, 3600 MW th , sodium fast reactor using a state-of-the-art low void effect reactor core design. The study focuses on the sodium boiling phenomenon and sodium reactivity feedback effect evolution during the accident with the reference subassembly (SA) design. Following this analysis, a study has been performed with a modified SA design. The boiling progression and phenomenology within the reference and the modified core have been compared, and the impact of the SA modification was described. [ABSTRACT FROM AUTHOR]

• New experiments for debris bed formation behavior in CDA of SFR are designed and conducted by releasing high-temperature particles. • The effect of coolant boiling is discussed detailly by qualitative observations and varying qualitative parametric analyses. • An empirical modeling extension function is developed to predict the flow-regime transitions underlying debris bed formation behavior with coolant boiling induced by discharged high-temperature debris. It is crucial to understand the Debris Bed Formation (DBF) behavior for the improved assessment of Core Disruptive Accident (CDA) of Sodium-cooled Fast Reactors (SFR). To elucidate the DBF flow-regime characteristics, a series of visualization simulated experiments were conducted at Sun Yat-Sen University by releasing solid particles into water pools. Because of different mechanisms of the inter-particle and fluid-particle interactions, four different flow regimes were identified along with four different final bed shapes found. Motivated to comprehensively understand the effect of coolant boiling on flow-regime transitions during the falling of high-temperature debris, new experiments are performed in present work by releasing high-temperature particles under varying parametric conditions. It is found that the boiling generated from falling high-temperature particles can effectively centralize the particle accumulation and diminish the particle-flow-induced pool convection, leading to the potential regime transitions. Based on the detailed experimental analyses and data attained, an extension function for evaluating the boiling effect caused by falling high-temperature particles is successfully developed to incorporate into our base model, whose applicability is examined to be restricted to the predictions under non-boiling condition. Knowledge from present work is useful for future development and verifications of SFR safety analysis models and codes in China. [ABSTRACT FROM AUTHOR]

The Versatile Test Reactor (VTR) currently under development is a 300 MWth sodium-cooled fast reactor (SFR) fueled with ternary metal alloy fuel, which aims to accelerate the testing of advanced nuclear fuels, materials, instrumentation, and sensors in high flux environments that are necessary to license the next generation of advanced reactor concepts. To support the VTR design process, uncertainties associated with the nuclear data has been propagated through the reactor core neutronics calculation to global parameters of interest, such as the core multiplication factor, kinetic parameters, and various reactivity feedback coefficients, following the sensitivity based uncertainty propagation approach. By folding the sensitivity coefficients, separately computed by the generalized perturbation theory code PERSENT and Monte Carlo code Serpent 2, with the variance–covariance matrices from COMMARA-2.0, we obtain the reaction-wise, isotope-wise, and overall uncertainties for each response of interest due to nuclear data uncertainty. With Serpent 2, the statistical error of the uncertainty is obtained by propagating the statistical error of the sensitivity coefficients through the same process using a newly developed uncertainty propagation method. From both codes, the overall top uncertainty contributors are found to be the cross section of Fe-56 elastic scattering, Na-23 elastic scattering, and U-238 inelastic scattering. The large contributions of the Fe-56 elastic scattering cross sections to global parameters are due to its relatively large relative uncertainty of 5–10% in nuclear data and the large volume of Fe-containing reflector assemblies in the fairly compact VTR core design. Both codes agreed well for the overall uncertainty estimates of all responses of interest, except the delayed neutron fraction, prompt neutron generation time, and the coolant density feedback coefficient, where Serpent 2 yielded a much larger value than PERSENT due to the large statistical error of sensitivity coefficients. The calculated uncertainties are also compared to those associated with other SFR cores. Another outcome of this study is a variance–covariance matrix of reactivity coefficients, which can be used in the subsequent uncertainty propagation to the system level to investigate the impact of identified uncertainties on system responses in the safety analysis. [ABSTRACT FROM AUTHOR]

• According to the application scenario of UUA, the concept of reactor with multi coolant flow and reactor active zone partition is proposed, which does not increase the core thermal load under the condition of high temperature rise. • For the special working conditions of reactor for UUA, the design scheme of special control rod structure is proposed. • MCNP-ORIGEN2 is used to build the reactor model and simulate the reactor criticality and burnup. The results show that the design scheme can meet the preset target of small reactor for unmanned underwater vehicle. In order to improve the endurance of unmanned underwater vehicle (UUV) and meet the needs of long-term underwater mission, the conceptual design of the 3.68 MWt long-life small reactor JCZER-20 is carried out in an UUV application scenario with a propulsive power of 850 kW. According to the design content, the concept of active zone partition design is put forward, and a new circulating scheme of coolant in the reactor primary circuit is given. The layout of control body in the reactor and the design of control system are carried out. The control rod structure of JCZER-20 adopts special design to deal with special conditions such as shaking during operation. The reactivity distribution scheme is given, and the reactor physical characteristics are analyzed for the design scheme. This scheme can lengthen coolant channels, effectively improve the total temperature rise of the coolant, but not increase the extra heat load. MCNP-ORIGEN2 simulation results show that the JCZER-20 reactor can operate at full power for more than 30 years without refueling. The reactor core reactivity can be effectively controlled by the control system of reactor. The core design meets preset design requirements. [ABSTRACT FROM AUTHOR]

• Wigner effect has been studied for decades because of its negative impact on the deployment and the operation of graphite-moderated reactors. • Recently, the possibility of storing energy through the purposeful creation of defects in crystalline materials via heavy particle bombardment was assessed. • The coupling of a Wigner effect-based energy storing concept with advanced reactor designs was investigated and the corresponding performance was evaluated. • Simulations confirmed that performance comparable with state-of-the-art batteries in terms of stored energy density can be achieved. In this work, an innovative energy storage concept based on the purposeful creation of defects in crystalline material by neutron irradiation is presented. Lattice defects are generated when heavy particles collide with the atoms in a crystal structure, i.e., if the incoming particles have enough energy, recoil atoms are displaced from their initial lattice sites. Most of the displaced atoms will eventually combine with nearby vacancies, but some of them will come to rest in non-ideal locations. The energy held by displaced atoms is called Wigner energy. Lattice defects can migrate and form clusters, and the Wigner energy can be released from these groupings if sufficient activation energy is provided. In the nuclear industry, this effect is well-known since it represented an issue for graphite-moderated reactors. This work presents the conceptual design of an engineering system that exploits this physical process to store the energy of neutrons in advanced reactor concepts. In the first part of the paper, the theoretical performance of an energy storage system based on the Wigner effect is described. Given the lattice properties and the compatibility with the harsh reactor environment, graphite was selected as the candidate material for the irradiation targets. Both experimental data and molecular dynamics simulations confirmed that this system can achieve performance comparable with state-of-the-art batteries in terms of stored energy density. In the second part of the paper, the engineering challenges of this innovative technology and the proposed solutions are described. After defining the optimal irradiation conditions, the different steps of the operation of the proposed energy system (from energy storing to energy harvesting) were defined. Finally, the integration of this concept with advanced reactor designs, i.e., a Sodium-cooled Fast Reactor and a Molten Salt-cooled Reactor, was investigated and the corresponding performance was evaluated. [ABSTRACT FROM AUTHOR]

• Droplets entrained by high-speed gas injection into a liquid pool was visualized. • Droplet diameter and velocity were measured simultaneously by image processing. • Complicated fluctuation of the jet interface greatly affected the droplet behavior. For the optimal design and safety evaluation of sodium-cooled fast reactors (SFRs), the risk estimation of tube failure in the steam generator (SG) via numerical analysis is very important. This requires the understand of the droplet entrainment behavior induced by high-speed steam injection from the heat transfer tube into the liquid sodium in the shell; however, there is a lack of detailed information such as droplet diameter and velocity due to measurement difficulties. Therefore, this study aims to understand the droplet entrainment caused by high-speed gas injection into a liquid pool and to obtain experimental data for developing proper analysis models. Air was injected into a water pool at various air superficial velocities and the droplets were visualized via the frame-straddling method, obtaining clear images of the droplet generation from the gas–liquid interface. The droplet entrainment was caused by the liquid phase stretching and tearing due to the gas jet shear stress. Based on the visualization results, the droplet diameter and velocity were simultaneously measured at various locations through image processing; the number of fast droplets decreased as the diameter increased. The droplet diameter increased along with the distance from the nozzle. At the center of the jet, the droplet velocity was higher near the nozzle and lower at larger distances. The droplets tended to be larger and slower near the jet interface than near the center of the jet. When the jet interface fluctuation became larger, the droplet diameter increased, while the velocity decreased. Therefore, to develop droplet entrainment analysis models for the risk evaluation of tube failures in SFR SGs, the effect of the jet interface fluctuation on the droplet behavior must be evaluated. [ABSTRACT FROM AUTHOR]

• The radiation energy source from the combusting sodium droplets is modeled. • The model is implemented into the multi-dimensional sodium fire analysis code. • The improved code simulates the sodium spray fire behavior reasonably. Heat radiation is one of dominant heat transfer processes during a sodium fire event which is a concern in sodium-cooled fast reactor plants. The multi-dimensional sodium fire analysis code AQUA-SF employs the 6-flux gas radiation model to simulate radiation heat transfer in atmospheric gas and to structural walls with the consideration of the radiation effects of suspended sodium by-products as aerosols. Here, this original model does not calculate radiation heat transfer from combusting droplets of sprayed sodium. All the combustion heat transports to atmospheric gas surrounding the combusting droplet except for the heat distributed to the droplet in this model. However, some of combusting heat should transport to structural walls directly by heat radiation in actuality. Due to this modeling, atmospheric temperature and pressure are overestimated and on the other hand, temperature of structural walls is underestimated. Therefore, this study aims to model radiation heat transfer from combusting droplets numerically. In the modeling, radiation energy transport on the combusting area around a sodium droplet is formulated considering emission, absorption and scattering as interaction with surrounding gas radiation. Radiation energy from the combusting droplets is added to the source term of the radiation transport equation in the 6-flux gas radiation model. Direct radiation heat transfer from combusting droplets can be simulated by this improved model. The improved model is tested through the verification analyses of single droplet combustion and the benchmark analysis on the upward sodium spray combustion experiment. The results of the test analyses show increase in heat transfer to the walls due to the droplet radiation. As the result, the sodium fire analysis becomes more reasonably by the improved model. [ABSTRACT FROM AUTHOR]

• We develop a model for melt relocation in a SFR fuel assembly with FAIDUS design. • The model is validated with the results of EAGLE ID1 and BLOKKER II test series. • Melt relocation characteristics for various FAIDUS design choices are investigated. • Dimensions of FAIDUS duct for a medium size SFR with MOX fuel are arrived at. The impediment of core disruption is realized through the provision of diverse and redundant safety features in the design complemented by the use of inherent passive mechanisms that respond to the accident condition by ensuring convenient dispersal of molten core material from the core in the early stage of the accident. In light of this, a promising design strategy Fuel Assembly Inner DUct Structure (FAIDUS) core concept is proposed to enhance the axial discharge capability of molten fuel from core for the exclusion of core wide melt pool formation. The sequence of event during the course of relocation in the FAIDUS is analysed employing a transient integrated thermal-hydraulic model. The effectual performance of various FAIDUS design options in mitigating the accident is evaluated. Based on this, the preferred design choice conforming the laid down criteria is selected in the present study. The simulation results suggest that FAIDUS design concept delivers the best alternative way for extenuating accident progress through a comprehensive melt relocation strategy of greater reliability in an intrinsic way. [ABSTRACT FROM AUTHOR]

The safety evaluation conducted for licensing a Sodium-cooled Fast Reactor (SFR) may require modeling transients with best-estimate calculation tools that must first be validated against real-world measurements. To provide the community with a valuable benchmarking opportunity for validating SFR analysis tools and methods, the International Atomic Energy Agency (IAEA) initiated a coordinated research project (CRP) in 2018 for the analysis of the Fast Flux Test Facility (FFTF) Loss of Flow Without Scram (LOFWOS) Test #13. The impact of nuclear data uncertainties on neutronics parameters was previously investigated based on the COMMARA-2.0 covariance matrix. Since the transient simulation results are very sensitive to certain reactivity coefficients, it was decided to employ rigorous uncertainty propagation methods to quantify the impact of nuclear data uncertainties on the best-estimate predication of FFTF LOFWOS Test #13. The DAKOTA code is used to propagate neutronic uncertainties through SAS4A/SASSYS-1 transient simulations, while taking into account spatial and reaction-wise correlations within these uncertainties. This study shows that the remaining discrepancies observed between the Argonne National Laboratory (ANL) best-estimate results and the experimental measurements can be partly explained by the uncertainty associated with Gas Expansion Module (GEM) worth, which contributes the majority of the overall nuclear data uncertainty on the output from the FFTF LOFWOS Test #13 transient simulation. This study also confirmed the importance of including spatial and reaction-wise correlations of nuclear data uncertainties on feedback coefficients in the uncertainty propagation to avoid under-estimating their impact during the transient simulations. [ABSTRACT FROM AUTHOR]

• New heat transfer correlation for sodium-to-sodium heat exchanger is proposed. • It is based on STELLA-1 experimental results. • Results are validated by STELLA-1 and JOYO experimental results. • Results show good agreement with the experimental results. Multiple studies have been conducted on the convective heat transfer of a liquid metal flow to present the correlations between the Nusselt and Péclet numbers. However, those correlations are considerably deviated from the experimental data obtained from steady-state tests of a sodium-to-sodium heat exchanger using the large-scale sodium thermal-hydraulic test facility STELLA-1. In this work, a new convective heat transfer correlation acquired through curve fitting with the STELLA-1 experimental results is proposed for a Péclet number range from 68 to 524. The newly proposed correlation was used in the SHXSA design code developed for designing a sodium-to-sodium heat exchanger by the Korea Atomic Energy Research Institute. The adequacy of the correlation was assessed by comparing the calculated results of the SHXSA code with the test results from STELLA-1. By using the proposed correlation, the average absolute deviations of the shell-side outlet temperature, tube-side outlet temperature, and heat transfer rate improved from 0.70% to 0.51%, 0.37% to 0.17%, and 1.61% to 0.82%, respectively. In the assessment with the experimental results of the JOYO sodium-to-sodium heat exchanger, the average absolute deviations of the shell-side outlet temperature, tube-side outlet temperature, and heat transfer rate were determined to be 0.27%, 0.54%, and 2.05%, respectively. The calculated results using the proposed correlation showed a satisfactory agreement with the experimental data from the STELLA-1 and JOYO facilities, and thus the proposed correlation is expected to be practically employed for designing a sodium-to-sodium heat exchanger. [ABSTRACT FROM AUTHOR]

• An introduction of the discrete adjoint sensitivity method. • A detailed implementation procedure of the discrete adjoint sensitivity method. • A verification of the correctness of the adjoint sensitivity method. • Analyses of sensitivities of temperature gradient in the hot plenum of an SFR. This paper presents a parameter sensitivity analysis on a thermal stratification (TS) model by using the discrete sensitivity method. The TS model was recently developed in our research group to efficiently predict the TS phenomenon in pool-type Sodium-cooled Fast Reactors. The fluid temperature gradient was considered as the figure of merit in the sensitivity analysis because it best characterizes the thermal stratification phenomenon. The sensitivities of the fluid temperature gradient with respect to four different parameters were investigated, including jet volumetric flow rate Q jet , jet temperature T jet , heat capacity of the ambient fluid C p , a m b , and static thermal conductivity of the ambient fluid k c , a m b. The sensitivity analysis was conducted through both the conventional forward sensitivity method and the advanced adjoint sensitivity method, which is more effective in cases where the number of outputs is small and the number of input parameters is large. The sensitivities obtained in this study suggested that perturbations in Q jet , C p , a m b , and k c , a m b could introduce either positive or negative changes to the temperature gradient, depending on the axial location and the elapsed time of the experiment. However, an increase in T jet always decreased the temperature gradient. Moreover, the impact of T jet on the maximum temperature gradient was several times higher than that of the other three parameters, which indicated that additional attention may need to be paid to the occurrence of thermal stratification in the sodium pool when the impinging jet has a large temperature change. This study also provides a step-by-step example for the application of the discrete adjoint sensitivity method to the time-dependent nonlinear systems. [ABSTRACT FROM AUTHOR]

• The RANS-based CFD are performed on the 37 and 61-pin wire-wrapped rod bundles. • The flow distribution is investigated in the subchannel with types and wire rotation. • The models on the flow split factor at the inner and edge subchannel was suggested. The wire-wrapped fuel bundle is commonly proposed in sodiumcooled fast reactors (SFRs) to enhance heat transfer of coolant in a compact triangular fuel array. In spite of its good performance, the wires around the fuel rods produce complex and periodic flow field, and it could eventually generate large uncertainties in design and licensing. In recent decades, the use of computational fluid dynamics (CFD) is gaining popularity to address this issue due to its capability to provide local flow and thermal information in large details. In this regard, CFD analyses have been performed here on the 37 and 61 pin SFR wire-wrapped fuel bundle experiment conducted at KAERI. The main purposes of this study are (1) to validate the CFD method in the SFR fuel bundle geometry, and (2) to investigate the local and the subchannel flow behaviors. First, the CFD simulations are compared with the experimental data and it is found that the applied models and the meshing schemes are good for prediction on the pressure drop and the mass flow rate for the given geometry. Afterward, the local flow data from the CFD is averaged for each subchannel in order to extract subchannel information. Then, the subchannel flow distribution characteristics are investigated and discussed using a flow split factor. Finally, this study proposes a model to predict the flow split factor in terms of subchannel types and wire rotation for the given wire-wrapped design. [ABSTRACT FROM AUTHOR]

• The frequency fluctuation coefficient is defined. • The relationship between the frequency fluctuation coefficient and the number of flow signal cycles is established. • A frequency fluctuation coefficient based bubbles detection method in sodium flow is proposed. • The proposed method is implemented on DSP-based hardware system in real-time. • The verification experiments are performed. The electromagnetic vortex flowmeter is a new type of instrument for detecting the steam generator leakage in sodium-cooled fast reactor for the operational safety of the entire nuclear power plant, and its performance is determined by the signal processing method. However, the existing signal processing methods pay more attention to the amplitude of the signal, and there are many factors that affect the amplitude of the signal, including the strength of the magnetic field, the flow rate of liquid sodium, and the magnification of the conditioning circuit. Therefore, a signal processing method of bubbles detection in sodium is studied from the perspective of frequency. It is found that the bubbles will cause a distorted waveform in the output signal of the OSOES 1 1 OSOES is short for output signal of the electromagnetic vortex flowmeter sensor. , thereby affecting the frequency calculation result of the OSOES. The flow signal cycle is used as the time unit to calculate the frequency of the OSOES, and the frequency fluctuation coefficient is defined. The relationship between the frequency fluctuation coefficient and the number of flow signal cycles is established to provide a strategy for studying how to highlight the effect of bubbles on the OSOES. A frequency fluctuation coefficient based bubbles detection method in sodium flow is proposed. It is implemented on DSP-based hardware system in real-time, and the verification experiments show that this method can detect the minimum water leak rate in the SG as low as 0.1 g/s. [ABSTRACT FROM AUTHOR]

• Implemented dynamic contact for reactivity control system within toolbox PIRAT. • Dynamic results are similar to quasi-static analysis with additional information. • Various excitation show the effects of nonlinearity in dynamic contact modeling. • Realistic modeling of earthquake shows little sensitivity to selected parameters. • Misalignment of guiding sleeves suggests extra research for ensuring reactor safety. Reactivity Control Systems (i.e., control rods and the associated drive mechanisms) are essential components to ensure the safe operation of nuclear reactors. Their design is particularly challenging during seismic events, due to the large deformations these impose on the core, which have the potential of hindering the insertion of the neutron absorbing material and delaying reactor shutdown. In order to assist the design process, the PIRAT toolbox is currently being developed for analytical based modeling of such systems. The present paper focuses on the development of the dynamic module DEBSE, and its application to Fast Reactor systems. Preliminary computations, based on simplified models and excitation that are progressively enriched to represent realistic situations, emphasize the role of dynamic effects. These computations show both situations where the static analysis is sufficient and insufficient to assess scram reliability due to their conservatism. While there are some issues with the numerical conditioning for this system, the results shows that for the most realistic seismic excitation available, safe shutdown is expected to occur with dynamic amplification occurring after the shutdown. This work also indicates possible causes to the ill-conditioning and what parameters are determined based on expert opinion that can be adjusted to better condition the equations of motion. Overall, this work shows the progress of PIRAT and shows examples of how it can be used for reactivity control system design. [ABSTRACT FROM AUTHOR]

• Model of natural convection of gas mixture with aerosol dynamics is developed. • Flow is dominated by a pseudo 2D roll-like structure. • Accumulation of droplets above 20 μm at the bottom is significant. • Pool temperature is a dominating factor of roof heat flux and aerosol concentration. • A correlation between aerosol concentration and pool temperature is provided. An argon gas cover layer is usually used in Sodium-cooled Fast Reactors (SFR) above the pool of sodium to prevent leakage of air into the vessel getting in contact with the liquid metal coolant. Heat transfer across the cover gas layer is of significant interest in terms of understanding the overall behaviour of the reactor and determining the thermal load of the roof slab, meanwhile, deposits of aerosol on to the structure surfaces across the cover gas region may hamper the operation of the penetrations and influence the coolant leakage detection system. In order to develop a comprehensive understanding of the thermal hydraulics in the cover gas region, a multi-physics model simulating the aerosol growth by condensation, aerosol transport and deposition, along with the natural convection of the gas mixture of argon and sodium vapour with thermal radiation across the semi-transparent mist has been developed. The model has been used for the simulation of a test facility for the study of cover gas dynamics and aerosol characteristics. The simulation results of the temperature profiles and aerosol concentrations under various conditions agree well with the experimental results. Under a normal operating condition of a sodium pool temperature of 550 °C, a roll-like structure is observed over the domain with a large circulation in one plane and several small vortices normal to it. Thermal radiation is found to be approximately 70% of the total heat flux through the roof. Gravitational settling is a significant factor in terms of aerosol dynamics which leads to an accumulation of droplets near the pool surface and the down flow region. An empirical correlation has been established based on the numerical and experimental data which shows that aerosol concentration increases with the sodium pool temperature. [ABSTRACT FROM AUTHOR]