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Investigations on the molten-pool sloshing behavior are of essential value for improving nuclear safety evaluation of Core Disruptive Accidents (CDA) that would be possibly encountered for Sodium-cooled Fast Reactors (SFR). This paper is aimed at synthesizing the knowledge from our recent studies on molten-pool sloshing behavior with solid particles conducted at the Sun Yat-sen University. To better visualize and clarify the mechanism and characteristics of sloshing induced by local Fuel-Coolant Interaction (FCI), experiments were performed with various parameters by injecting nitrogen gas into a 2-dimensional liquid pool with accumulated solid particles. It was confirmed that under different particle-bed conditions, three representative flow regimes (i.e. the bubble-impulsion dominant, transitional and bed-inertia dominant regimes) are identifiable. Aimed at predicting the regime transitions during sloshing process, a predictive empirical model along with a regime map was proposed on the basis of experiments using single-sized spherical solid particles, and then was extended for covering more complex particle conditions (e.g. non-spherical, mixed-sized and mixed-density spherical particle conditions). To obtain more comprehensive understandings and verify the applicability and reliability of the predictive model under more realistic conditions (e.g. large-scale 3-dimensional condition), further experimental and modeling studies are also being prepared under other more complicated actual conditions.

Among Generation IV reactors, the sodium-cooled fast reactor (SFR) is attracting attention as a system having great potential for commercial use. Gas entrainment is a thermal-hydraulic issue related to the safety problem of the reactor core in the SFR. Typically, a dipped plate or baffles are installed under the free surface to suppress gas entrainment. However, these approaches can cause gas entrainment in other locations and require many trial-and-error and verifications. In this study, a new strategy using magnetohydrodynamics to suppress gas entrainment in the SFR is proposed. In a counter-flow model, a judgment criterion of gas entrainment occurrence was developed for both water and liquid metal. Moreover, the gas entrainment can be completely suppressed by applying a magnetic field.

Sodium-cooled fast reactor (SFR) is the preferred technology of the generation-IV fast neutron reactor, and its core body mainly uses nuclear-grade 316 stainless steel. In order to prolong the design life of SFRs to 60 years and more, it is necessary to summarize and analyze the anti-corrosion effect of nuclear grade 316 stainless steel in high temperature sodium environment. The research on sodium corrosion of nuclear grade 316 stainless steel is mainly composed of several important factors, including the microstructure of stainless steel (ferrite layer, degradation layer, etc.), the trace chemical elements of stainless steel (Cr, Ni and Mo, etc) and liquid impurity elements in sodium (O, C and N, etc), carburization and mechanical properties of stainless steel, etc. Through summarizing and constructing the sodium corrosion rate equations of nuclear grade 316 stainless steel, the stainless steel loss of thickness can be predicted. By analyzing the effects of temperature, oxygen content in sodium and velocity of sodium on corrosion rate, the basis for establishing integrity evaluation standard of SFR core components with sodium corrosion is provided.

In a postulated severe accidental condition of sodium-cooled fast reactor (SFR), the eutectic reaction between boron carbide (B4C) as control rod element and stainless steel (SS) as control rod cladding or related structure materials may take place. Thus, the kinetic behavior of the B4C-SS eutectic reaction is one of the important phenomena to be considered when evaluating the core disruptive accidents in SFR. In this study, for the first step to obtain the fundamental information on kinetic feature of the B4C-SS eutectic reaction and compare the pervious findings, the thermal analysis using the pellet samples of B4C and Type 316L SS as a different experimental approach was performed up to 1773 K at different heating rates of 2.5-10 K min-1. The differential thermal analysis (DTA) endothermic peaks for the B4C-SS eutectic reaction appeared from 1483K to 1534K and systematically shifted to higher temperatures with increasing heating rate. Based on this kinetic feature, apparent activation energy and pre-exponential factor for the B4C-SS eutectic reaction were determined by Kissinger method. It was found that the kinetic parameters obtained by thermal analysis were comparable to the literature values of the thinning experiment at high temperatures. In addition, the microstructure and element distribution formed in the interdiffusion layer composed of the B4C-SS system were analyzed by the electron probe microanalyzer (EPMA), which can provide key validation data on elemental interdiffusion behavior in the early stage of the eutectic reaction for reflecting the reaction kinetic modeling.

Gas entrainment (GE) from cover gas, which is an inert gas to cover sodium coolant in the reactor vessel, is one of the key issues for Sodium-cooled fast reactors (SFRs) design to prevent unexpected effects to core reactivity. In this research series, evaluation method has been investigated for surface dimple depth growth of unstable drifting vortex dimples on the liquid surface in the reactor vessel. By using a computational fluid dynamics (CFD) code, analyses have been conducted to estimate the drifting vortex on water experiments in a circulating water tunnel. The unstable drifting flow vortexes on the water surface were generated as wake vortexes behind a plate obstacle. Downward flow velocity was induced by the bottom slit’s flow passing along the flow channel. In the previous study, the initial conditions of the gas entrainment were evaluated based on existing non-dimensional numbers method by using the STREAM-VIEWER code. However, the CFD predication accuracy of the detailed flow field itself was not clear especially for vortex frequency in the wake flow and detailed velocity profiles in the flow channel. In this study, to clarify the accuracy of CFD analysis, Strouhal numbers of vortex frequency and detailed flow velocity profiles were compared with experimental data which were measured by Particle Image Velocimetry (PIV) method. As the results, the Strouhal numbers of the vortex frequency behind the plate obstacle reasonably agreed with the experimental data.

Robust manufacturing technology for sodium-cooled fast reactor (SFR) components required development of various innovative welding technologies backed by thorough research. Construction of the 500 MWe Prototype Fast Breeder Reactor at Kalpakkam was preceded by extensive technology development which included planning and implementation of R&D on welding consumables and procedures and also technologies for fabrication of components with stringent specifications. To take on the challenges and establish robust manufacturing techniques for fabrication of all the structural components of the SFR, a close interaction among design, materials and non-destructive evaluation engineers, materials and welding consumable manufactures, and the fabrication industries had to be facilitated. Fabrication of SFR reactor vessels, steam generators, fuel sub-assemblies, etc., requires extensive welding. Activated tungsten inert gas (A-TIG) welding process together with activated flux was developed in-house. The same was successfully implemented on various SFR components. Online monitoring and control during welding was developed based on a computational intelligence methodology and deployed for obtaining near defect-free welded components. Plasma transferred arc welding process was used for deposition of a nickel-base alloy for hardfacing of SFR components, with improved resistance to high-temperature wear, especially galling of mating surfaces in liquid sodium environment. In order to substantially reduce carbon footprint, India has initiated a national mission programme for design, development and establishment of a 800 MWe Advanced Ultra-supercritical (AUSC) power plant having steam parameters of 710oC/ 720oC/ 310 bar. Success of this program heavily banks on development of materials and fabrication technologies that can sustain the extreme in-service conditions. Two important India-specific high-temperature tube materials were developed for the purpose, namely, 304HCu austenitic stainless steel (304HCu SS) and modified Alloy 617M. In addition, large cylindrical forgings of Alloy 617M of up to 800 mm diameter for turbine rotors were indigenously developed. R&D activities included development of welding consumables and welding procedure for TIG welding of tubes, including the 304HCu SS/Alloy 617M dissimilar joint. Based on the tensile and creep tests results, welding consumables for joining the tubes and their dissimilar joints were also finalized. Technology development also included dissimilar metal welding between Alloy 617M and 10Cr-steel forgings for turbine rotors. The dissimilar metal weld between 400 mm Alloy 617M and 10Cr-steel were produced by directly welding the two forgings using Alloy 617 filler wire using narrow-gap TIG welding process. Similarly, procedure for NG welding of superheater headers was also demonstrated. This paper discusses the Indian efforts toward development and deployment of welding technologies for fabrication of components for the Indian SFR-based nuclear power plant and the AUSC thermal power plant.

SS316Ti is one of the widely used materials for bellows in sodium cooled fast reactor systems. Conventional design codes for bellows do not address the high-temperature failure modes. In this investigation, a methodology has been proposed for the high-temperature design of the bellows. The bellows were preliminarily designed by standards of EJMA for room temperature application. Subsequently, the detailed visco-plastic finite element analysis of the bellows has been carried out to estimate the high-temperature low-cycle fatigue (LCF) and creep-fatigue (CFI) life. The material parameters required for the visco-plastic constitutive model were generated and validated for fatigue specimens. The LCF and CFI life of the bellows based on analysis were 83 and 74 cycles, respectively. The proposed methodology has been validated by high-temperature testing of the bellows. The LCF and CFI lives of the bellows obtained experimentally were found to be 281 and 237 cycles, respectively. The life arrived based on analysis was found to be conservative compared to corresponding life by testing. The safety margin available over failure based on testing of the bellows is 3.4 under LCF and 3.2 for CFI loading.

This paper presents results of calculation studies of the viability of coated particles in the conditions of the reactor core on fast neutrons with sodium cooling, justifying the development of the concept of the reactor BN with microspherical fuel. Traditional rod fuel assemblies with pellet MOX fuel in the core of a fast sodium reactor are directly replaced by fuel assemblies with micro-spherical mixed (U,Pu)C-fuel. Due to the fact that the micro-spherical (U, Pu)C fuel has a developed heat removal surface and that the design solution for the fuel assembly with coated particles is horizontal cooling of the microspherical fuel, the core has additional possibilities of increasing inherent (passive) safety and improve the competitiveness of BN type of reactors. It is obvious from obtained results that the microspherical (U, Pu)C fuel is limited with the maximal burn-up depth of ∼11% of heavy atoms in conditions of the sodium-cooled fast reactor core at the conservative approach; it gives the possibility of reaching stated thermal-hydraulic and neutron-physical characteristics. Such a tolerant fuel makes it less likely that fission products will enter the primary circuit in case of accidents with loss of coolant and the introduction of positive reactivity, since the coating of microspherical fuel withstands higher temperatures than the steel shell of traditional rod-type fuel elements.

This paper suggests a fracture simulation method for SFR metallic fuel pin under accident condition. Two major failure mechanisms - creep damage and eutectic penetration - are implemented in the suggested method. To simulate damaged element, stress-reduction concept to reduce stiffness of the damaged element is applied. Using the proposed method, the failure size of cladding can be predicted in addition to the failure time and failure site. To verify the suggested method, Whole-pin furnace (WPF) test and TREAT-M test conducted at Argonne National Laboratory (ANL) are simulated. In all cases, predicted results and experimental results are overall in good agreement. Based on the simulation result, the effect of eutectic-penetration depth representing failure behavior on failure size is studied.

In order to improve the safety of nuclear power plants, it is necessary to make sure measures against their severe accidents. Especially, in the case of a sodium-cooled fast reactor, there is a possibility that molten core material would be discharged through control rod guide tubes into the inlet coolant plenums beneath the rector cores in the event of a core disruptive accident (CDA). It is important to ensure in-vessel retention that keeps and confines damaged core material in the reactor vessel even if the CDA occurs. In this study, effective cooling of the melt in coolant was confirmed by comparing the experiment and analysis. CDA scenario initiated by a unprotected loss of flow condition , which is a typical cause of core damage, is generally categorized into four phases according to the progression of core-disruptive status, which are the initiating, early-discharge, material-relocation and heat-removal phases for the latest design in Japan. During the material-relocation phase, the molten core material flows down mainly through the control rod guide tube and is discharged into the inlet coolant plenum below the bottom of the core. The discharged molten core material collides with the bottom plate of the inlet plenum. Clarification of the accumulation behavior of molten core material with such a collision on the bottom plate is important to reduce uncertainties in the safety assessment of CDA. In present study, in order to make clear behavior of core melt materials during the CDAs of sodium-cooled fast reactors, analysis was conducted using the SIMMER-III code for melt discharge simulation experiments by Imaizumi et al. in which low-melting-point alloy was discharged into a shallow water pool. As the result, temperature and pressure behaviors during the discharge almost coincided between the analysis and the experiment. Therefore, it can be concluded that the validity of the analysis cell system model was confirmed.

A sodium-cooled fast reactor has been designed to attain a high burn-up core in commercialized fast reactor cycle systems. The sodium-cooled fast reactor adopts a wire spacer between fuel pins. The wire spacer performs functions of securing the coolant channel and mixing between subchannels. In high burn-up fuel subassemblies, the fuel pin deformation due to swelling and thermal bowing may decrease the local flow velocity in the subassembly and influence the heat removal capability. Therefore, understanding the flow field in a wire-wrapped pin bundle is important. This study performed particle image velocimetry (PIV) measurements using a wire-wrapped three-pin bundle water model to grasp the flow field in the subchannel under conditions, including the laminar to turbulent regions. The PIV results confirmed that the normalized flow velocity near the wrapping wire in the low Re number condition was relatively decreased compared to that in the high Re number condition. Meanwhile, in the region away from the wrapping wire, the maximum flow velocity was increased by decreasing the Re number. Accordingly, the PIV measurements using the three-pin bundle geometry without the wrapping wire were also conducted to understand the effect of the wrapping wires on the flow field in the subchannel. The results confirmed that the mixing due to the wrapping wire occurred, even in the laminar condition. These experimental results are useful not only for understanding the pin bundle thermal hydraulics, but also for the code validation.

SAS4A is presently the worldwide standard computer code for simulation of the initiating phase of the Core Disruptive Accident (CDA) in MOX-fueled Sodium-cooled Fast Reactors (SFR). In order to imp...

Thermal stratification (TS) is a thermal-fluid phenomenon that can introduce large uncertainties to nuclear reactor safety. The stratified layers caused by TS can lead to temperature oscillations i...

Methods of determining the efficiency of the system that controls the seal-tightness of fuel-rod cladding and localizes FA with leaky fuel rods in a fast reactor are examined. It is shown that the design procedure has significant limitations. A procedure for numerical modeling of the transport of delayed-neutron precursors was developed to take account of the special features of liquid-metal coolant flow. A special computational module FV-BN was developed within the framework of the FlowVision software package. The computational results obtained for the concentration distribution of delayed-neutron precursors are transferred into the deterministic transport code TORT in order to obtain the spatial-energy distribution of the neutron flux density in a three-dimensional geometry. The procedure was verified on full-scale reactor problems by simulating the flow-through parts of the upper mixing chamber of the fast reactor.

An annular linear induction electromagnetic pump (ALIP) which has a developed pressure of 0.76 bar and a flow rate of 100 L/min is designed to analysis end effect which is main problem to use ALIP in thermohydraulic system of the prototype generation-IV sodium-cooled fast reactor (PGSFR). Because there is no moving part which is directly in contact with the liquid, such as the impeller of a mechanical pump, an ALIP is one of the best options for transporting sodium, considering the high temperature and reactivity of liquid sodium. For the analysis of an ALIP, some of the most important characteristics are the electromagnetic properties such as the magnetic field, current density, and the Lorentz force. These electromagnetic properties not only affect the performance of an ALIP, but they additionally influence the end effect. The end effect is caused by distortion to the electromagnetic field at both ends of an ALIP, influencing both the flow stability and developed pressure. The electromagnetic field distribution in an ALIP is analyzed in this study by solving Maxwell's equations and using numerical analysis.

The Floating Absorber for Safety at Transient (FAST) is a safety device used in the innovative Sodium-cooled Fast Reactor (iSFR). The FAST insert negative reactivity under transient or accident conditions. However, behavior of the FAST is still unclear under transient conditions. Therefore, the existing Floating Absorber for Safety at Transient Analysis Code (FASTAC) is improved to analyze the FAST movement by considering the reactivity and temperature distribution within the reactor core. The current FAST system is simulated under a single control rod withdrawal accident condition. In this investigation, the reactor thermal power does not return to its initial thermal power even if the FAST inserts negative reactivity. Only a 9 K of coolant temperature margin, in the hottest fuel assembly at EOL, can lead to unnecessary insertion of the negative reactivity. On the other hand, the FASTs cannot contribute to controlling the reactivity when normalized radial power is less than 0.889 at BOL and 0.972 at EOL. These simulation results suggest that the current FAST design needs to be optimized depending on its installed location. Meanwhile, the FAST system keeps the fuel, cladding and coolant temperatures below their limit temperatures with given conditions.

Core-wide temperature distribution in sodium-cooled fast reactor plays a key role in its decay heat removal process, however the prediction for temperature distribution is quite complex due to the conjugate heat transfer between the assembly flow and the inter-wrapper flow. Hybrid medium model has been proposed for conjugate heat transfer modeling in the core. The core is modeled with a Realistic modeled inter-wrapper flow and hybrid medium modeled assembly flow. To validate present model, simulations for a three-assembly model were performed with Realistic modeling, traditional porous medium model and hybrid medium model, respectively. The influences of Uniform/Non-Uniform power distribution among assemblies and the Peclet number within the assembly flow have been considered. Compared to traditional porous medium model, present model shows a better agreement with in Realistic modeling prediction of the temperature distribution and the radial heat transfer between the inter-wrapper flow and the assembly flow. Keywords: Porous medium model, Hybrid medium model, Conjugate heat transfer, Inter-wrapper flow, Sodium-cooled fast reactor

As the refueling of a sodium-cooled fast reactor is conducted by rotating part of the reactor head without opening it, the monitoring of existing obstacles that can disturb the rotation of the reactor head is one of the most important issues. This paper deals with the ultrasonic ranging technique that directly monitors the existence of possible obstacles located in a lateral gap between the upper internal structure and the reactor core in a prototype generation IV sodium-cooled fast reactor (PGSFR). A 10 m long plate-type ultrasonic waveguide sensor, whose feasibility has been successfully demonstrated through preliminary tests, was employed for the ultrasonic ranging technique. The design of the sensor's wave radiating section was modified to improve the radiation performance, and the radiated field was investigated through beam profile measurements. A test facility simulating the lower part of the upper internal structure and the upper part of the reactor core with the same shapes and sizes as those in the PGSFR was newly constructed. Several under-water performance tests were then carried out at room temperature to investigate the applicability of the developed ranging technique using the plate-type ultrasonic waveguide sensor with the actual geometry of the PGSFR's internal structures. Keywords: Plate-type ultrasonic waveguide sensor, Ultrasonic ranging, Lamb wave, Sodium-cooled fast reactor

The authors are developing the design concept of the pool-type sodium-cooled fast reactor (SFR) that addresses Japan’s specific siting conditions such as earthquakes and meets safety design criteria (SDC) and safety design guidelines (SDG) for Generation IV SFRs. The development of this concept will broaden not only options for reactor types in Japan but also the range and depth of international cooperation. A design concept of 1,500 MWt (650 MWe) class pool-type SFR was thought up by applying design technology obtained from the design of advanced loop-type SFR, named JSFR, equipped with safety measures that reflect results from the feasibility study on commercialized fast reactor cycle systems and fast reactor cycle technology development, improved maintainability and repairability, and lessons learned from the Fukushima Daiichi Nuclear Power Plants accident. The design concepts of a reactor vessel (RV) and its internal structures have been investigated whether they could withstand severe seismic conditions in Japan and thermal loads. The design adopted enhanced RV support structure, enhanced conical-shaped core support structure, a thickened knuckle part of the RV, and a flat plenum separator with ribs. A three-dimensional steady-state thermal hydraulic analysis of the RV revealed that the temperature difference of the upper and lower surfaces of the flat plenum separator could be effectively reduced by installing layers of thermal insulation plates. The authors have also conducted a transient analysis of loss of flow type anticipated transients without scram events to evaluate the feasibility of a self-actuated shutdown system. Moreover, the configuration of the decay heat removal system has been investigated considering sufficient utilization of natural circulation capability of sodium coolant, heat removal capacity of each cooling system, conformance with design requirements, and recommendations of SDC and SDG such as diversity and redundancy of components.

If pressurized water or its vapor leaks from a ruptured heat transfer tube in a steam generator of a sodium-cooled fast reactor, a high-velocity, high-temperature, and corrosive jet with sodium-water chemical reaction may cause tube failure propagation. In this study, an analytical method was developed to predict the occurrence of tube failure propagation by overheating rupture. This method consists of the elemental analytical models for a sodium-side temperature distribution formed by a reacting jet, water-side thermal hydraulics, heat transfer between a fluid and a tube, and tube failure by internal pressure. To evaluate the tube failure propagation in a short computation time, these models are based on the experimental data, semi-theoretical correlations, or one-dimensional equations. Applicability of the proposed method was investigated through the numerical analysis of an experiment on water vapor discharging into the liquid sodium. This analysis demonstrated that the method could predict the occurrence of overheating rupture and provide conservative results. While the proposed method is useful for high-speed computations, this method evaluates a high temperature region with a large conservativeness in some cases. To improve this conservativeness, a Lagrangian particle model for the reacting jet was also developed as an alternative method. The numerical analysis by this model showed that the discharged gaseous particles spread with particle-particle and particle-tube interactions.

Many components in the assembly section of Sodium-cooled Fast Reactor are made of good corrosion-resistant 316 LN Stainless Steel material. To avoid self-welding of the components with the coolant sodium at elevated temperature, hardfacing is inevitable. Ni-based colmonoy-5 is used for hardfacing due to its lower dose rate by Plasma Transferred Arc process due to its low dilution. Since Ni–Cr–B–Si alloy becomes very fluidic while depositing, the major height of the weld overlay rests inside the groove. Hardfacing is also done over the plain surface where grooving is not possible. Therefore, grooved and un-grooved hardfaced specimens were prepared at different travel speeds. Fe content at every 100 μm of the weld overlay was studied by Energy Dispersive Spectroscopy and also the micro hardness was determined at those locations. A correlation between iron dilution from the base metal and the micro hardness was established. Therefore, if the Fe content of the weld overlay is known, the hardness at that location can be obtained using the correlation and vice-versa. A new correlation between micro hardness and dilution coefficient is obtained at different locations. A comparative study between those specimens is carried out to recommend the optimum travel speed for lower dilution. Keywords: Dilution, Hardness, Ni alloy, Austenitic stainless steel, Plasma transferred arc hardfacing (PTAH), Travel speed

Fully natural circulation decay heat removal systems (DHRSs) are adopted for sodium fast reactors, which is a passive safety feature without any electrical pumps. It is needed to grasp the thermal-hydraulic phenomena in the reactor vessel and evaluate the coolability of the core under the natural circulation not only for the normal operating condition but also for severe accident conditions. In this paper, the numerical results of the preliminary analysis for the sodium experimental condition with the PLANDTL-2, in which the core and the upper plenum with a dipped-type direct heat exchanger (DHX) were modeled, are discussed to establish an appropriate numerical models for the reactor core including the gap region among the subassemblies and the DHX. The transient analysis simulating the reactor scram reveals that the 3-dimensional large scale flow structure is developed through the gaps in the whole of the core area during the reactor scram. The steady-state analysis coinciding Richardson number between the PLANDTL-2 and the reactor operation condition reveals that the hot spot and cold spot appear depending on the location of the DHX, which is caused by the complex thermal-hydraulic phenomena driven by the natural circulation. From these preliminary analyses, the characteristics of the thermal-hydraulics behavior in the PLANDTL-2 to be focused are extracted.

International audience; The Euratom Horizon-2020 project European sodium fast reactor safety measures assessment and research tools (ESFR-SMART) aims to enhance the safety and performance of the European sodium-cooled fast reactor (ESFR) considering safety objectives envisaged for generation-IV reactors and the update of European and international safety frameworks, taking into account the Fukushima accident. Further, the project aims to support the development and validation of the computational tools for the situations to be considered at each defense-in-depth level in order to support the safety assessments using data produced in the project as well as selected legacy data. Within this activity, the focus is on the assessment and when needed further development of computer codes for the analysis of the sodium thermal-hydraulics phenomena in sodium cooled fast reactor (SFR) subassemblies under operational and accidental conditions including sodium boiling and transitional convection from forced to natural/mixed convection. In support of this activity, a review is performed for the sodium boiling models used in the codes participating in benchmark activities within the project. The objective of this paper is to summarize the result of the review, which encompasses both the phenomenological and mathematical models implemented in the codes. In particular, the review addresses the physical bases of sodium boiling models, the analytical models in connection with the numerical implementations in the codes, and the geometry representations ranging from a one-dimensional single channel and finite volume to full three-dimensional computational fluid dynamics (CFD) and porous media representations. Finally, the needs for further developments are discussed.

Hexcan seal welds in fuel subassemblies of prototype fast breeder reactor (PFBR) pose serious problem during ultrasonic inspection for defects due to lower thickness involved and its complex structure. This paper presents detection of defects and identification of its localization across the thickness in fuel subassembly hexcan seal welds using high frequency guided (plate) waves (HFGW). The generation and propagation characteristic of the optimized guided wave mode is extensively studied by using 2D finite element method (FEM). The energy transfer between the hexcan sheath surfaces results in a better sensitivity for detection of small defects on both surfaces. The study demonstrates that the through-thickness location of the defects can be accurately determined by using the frequency content rather than time-of-flight information.

The Leak Before Break (LBB) technology plays a vital role in the safety evaluation of Sodium-cooled Fast Reactor (SFR). In this study, the leak rate considered in LBB assessment of SFR was predicted and analyzed. Based on the models of fluid resistance in cracks, two-dimensional numerical simulation for leakage process of uniform cross-section axial cracks were conducted with Computational Fluid Dynamics (CFD) to obtain recommended values of crack empirical parameters, including entrance resistance coefficient, effective parameters and friction factors. With consideration of the complex crack morphologies, the leak rate prediction models with a wide range of parameters were established for liquid sodium, and corresponding computation code (LAPIS) was developed. The proposed models were verified and validated by seepage model, numerical results and experimental data with deviations within 1%, 35% and 10%, respectively. The effects of the crack morphology and thermal-hydraulic parameters on the leak rate for tight and wide cracks were studied with the LAPIS. The leak rate increases with Crack Opening Displacement (COD), crack length, ratio of entrance to exit area, entrance resistance coefficient, stagnation pressure and decreases with temperature. This work could provide theoretical and technical support for the study of leak rate considered in LBB assessment of SFR.

The steam generator in a nuclear power plant is a type of heat exchanger in which heat transfer occurs from the hot fluid in multiple channels to the cold fluid. Therefore, a uniform flow over multiple channels is necessary to improve heat exchanger efficiency. The study aims at experimentally investigating the improvement of flow uniformity by the perforated plate in the heat exchanger used for a sodium-cooled fast reactor stream generator. A 1/4-scale experimental model for one heat exchanger unit with 33 × 66 channels was manufactured. The working fluid was water. A perforated plate was systematically designed using numerical simulations to improve the flow uniformity over the 33 × 66 channels. As a result, the flow uniformity greatly improved at a slight cost of pressure drop. To validate the numerical results, planar particle image velocimetry measurements were performed on the selected planes in the inlet and outlet headers. The experimental velocity profiles near the exits of the channels were compared with numerical simulation data. The experimental profiles agreed with the numerical data well. Both the numerical simulation and the experimental results showed a slight increase in pressure drop, despite significant improvement in the flow uniformity.

As a critical component for pool-type Sodium cooled Fast Reactor (SFR), the main vessel plays a major role in holding the entire sodium in the pool. While the Reactor Vessel Cooling System (RVCS) would contribute to the cooling of the main vessel. This paper develops a one-dimensional code and the numerical analysis has been conducted to evaluate the thermal-hydraulic characteristics for RVCS for China Experimental sodium-cooled Fast Reactor (CEFR). Firstly, the code is applied to the analysis on the steady characteristics of RVCS and the results are compared with the design value. The simulation results show that the main vessel is cooled effectively and the comparison results are in good agreement with the design value. Then, in order to study the transient characteristics of RVCS, the code is coupled with the Transient Thermal-Hydraulic Analysis Code for Sodium-cooled fast reactors (THACS). The simulation finds that a reverse flow in RVCS shows up and RVCS could take the decay heat out effectively, which contributes a lot to the core cooling. The code evaluated the thermal-hydraulic characteristics as well as the decay heat removal capacity of RVCS. The results are very promising but further experimental data is still needed to validate the code.

To ensure the safe operation of nuclear power plant, employees must complete nuclear power plant-related safety training. College students who are majored in nuclear engineering in China cannot deeply participate in nuclear power plant work during the college period. So, they do not have a deep understanding of the theoretical concepts of nuclear power systems in class. For research and teaching work on Sodium-cooled fast reactor (SFR), the Department of Engineering Physics at Tsinghua University has developed the sodium-cooled fast reactor nuclear power plant simulation software (ECFR) with Neoswise Science & Technology Co. Ltd. ECFR is the dynamic process model software of the SFR nuclear power plant which is based on VPOWER platform. This paper presents the application of ECFR in teaching. Experimental courses have been conducted for Tsinghua students. Students can observe the main system structure of the power plant and perform accident analysis through ECFR. Those experiments can help students grasp the latest nuclear power technology better and motivate them to study effectively. Furthermore, the future improvement direction of ECFR is also discussed, which includes the improvement measures of human-computer interaction and system design.

In the design study of an advanced sodium-cooled fast reactor (Advanced-SFR) in Japan Atomic Energy Agency (JAEA), the use of a specific fuel assembly (FA) with an inner duct structure called FAIDUS has been investigated to enhance safety of Advanced-SFR against a core disruptive accident. Due to the asymmetric layout of fuel rods by the inner duct, it is necessary to estimate the temperature distribution to confirm feasibility of FAIDUS. In JAEA, an in-house subchannel analysis code named ASFRE has been developed as a FA design tool. For the typical FAs, the numerical results of ASFRE were validated by comparisons with experimental data. As for the FAIDUS, confirmation of validity of the numerical results by ASFRE was not enough because the reference data on the thermal hydraulics in FAIDUS have not been obtained by the mock-up experiment, yet. In this paper, therefore, the code-to-code comparisons with numerical results of ASFRE and those of an in-house CFD code named SPIRAL was conducted to make further discussion on applicability of ASFRE to the thermal-hydraulics analysis of FAIDUS. The applicability of ASFRE was indicated through the confirmation of the consistency of mechanism of the specific temperature and velocity distributions appearing around the inner duct between the numerical results by ASFRE and those by SPIRAL.

An MHD pump with a developed pressure of 4 bar and flow rate of 20 kg/s was designed using electrical equivalent circuit method, which is typically used in analyzing linear induction machines. The experiment is conducted at various input power and frequency conditions. The characteristics of MHD pump, such as the mass flow rate and developed pressure, are analyzed at varying input power and frequency. The main design variables of the pump are theoretically calculated for a sodium-cooled fast reactor thermal hydraulic experimental loop. P–Q characteristics experiments of the MHD pump are conducted at a flow rate of 100 L/min and developed pressure of 0.76 bar to compare theoretical and experimental results.

An investigation on the nuclear transmutation of elemental long-lived fission product (LLFP) in a fast reactor is being conducted focusing on the I-129 LLFP (half-life 15.7 million years) to reduce the environmental burden. The LLFP assembly is loaded into the radial blanket region of a Japanese MONJU class sodium-cooled fast reactor (710 MWth, 148 days/cycle). The iodine element containing I-129 LLFP (without isotope separation) is mixed with YD2 and/or YH2 moderator material to enhance the nuclear transmutation rate. We studied the optimal moderator volume fraction to maximize the transmutation rate (TR, %/year) and the support factor (SF is defined as the ratio of transmuted to produced LLFP). We also investigated the effect of LLFP assembly loading position in the radial blanket and the severe power peak appeared at the fuel assembly adjacent to the LLFP assembly.

The SAS4A safety analysis code, originally developed for the analysis of postulated severe accidents in oxide fuel sodium-cooled fast reactors (SFRs), has been significantly extended to all...

A non-classical porous body model has been adopted to predict the thermal hydraulics within a totally blocked prototype spent fuel subassembly of a sodium cooled fast reactor. The heat transfer within the subassembly is by natural convection of sodium and the ultimate heat sink is inter-wrapper flow. The decay heat is considered as a source term in the energy equation. Governing equations for 3-D, steady state, porous-body based natural convection have been solved by using the commercial CFD code ANSYS WORKBENCH 15.0. The present study considers a 217 pin wire wrap bundle of spent fuel subassembly, for an axial length of five helical pitches. The magnitudes of maximum sodium temperature and natural convection velocity inside the blocked subassembly have been obtained from the model. Through parametric studies, the permissible decay power (i) that avoids boiling of sodium trapped inside the subassembly (450 kW) and (ii) that restricts the peak clad temperature less than 823 K (100 kW), have been obtained. Further, an equivalent conduction model for the natural convection has been developed which can be used for quick estimate of sodium temperature at various values of decay power. It is seen that the conductivity value of sodium in the equivalent conduction model varies as keq(q′′′) = kNa + C×(q′′′)n where the values of C and n are found to be 1 and 0.24. From the point of view of safety, subassembly outlet sodium temperature monitoring has to be initiated at this power level itself, during power raising operation in a fast reactor. The permissible decay power from clad safety point of view is reached 6 h after reactor shut-down. This suggests that fuel handling operation can be initiated ideally after this time.

In the event of Core Disruptive Accident (CDA) condition in a sodium cooled fast reactor (SFR), the environmental source term (i.e. quantity of radioactivity released into the environment surrounding to the reactor) is determined by the activity released from the containment building which in-turn depend on the concentration of radioactive aerosols remain suspended in the containment. The aerosol concentration depletes with progress of time inside the containment. Various models were used to predict the time evolution of aerosol concentration inside the containment. It is observed from the literature that there exists significant difference in the model prediction with simulated experimental condition. One of the reasons is the aerosol characteristics used as an input in the models is not accurate. In particular, the sodium aerosol characteristics (generated by the combustion of sodium) get changed both by physical and chemical processes. The other aerosols like fission products, fuel and structural materials form aggregates with sodium aerosols and their morphological properties are also get changed with progress of time. Hence an experimental study is carried out to determine the morphological properties viz. particle diameters, particle density, effective density, shape factors of sodium combustion aerosols, fission product aerosols (Cerium Oxide and Strontium oxide), structural aerosols (D9 steel) and mixed aerosols of sodium and fission products; and reported in this paper. Since sodium aerosols are hygroscopic, experiments are conducted at 3 different humidity conditions. In all the cases, the aerosol density is found to be lesser than the material density, and it changes with time towards the steady state. In case of sodium aerosols for mass concentration of 2 g/m3, the average particle densities and the standard relative uncertainties arrived at 75%, 55% and 30% RH are 1.67 ± 0.41, 1.43 ± 0.36 and 1.2 ± 0.31 respectively. Sodium aerosols are mostly spherical at humidities of 55% and 75% RH, where as they are deviated from spherical at 30% RH. Particle densities of D9 steel, CeO2 and SrO2 aerosols are 3.92 ± 0.58, 4.27 ± 0.54 and 3.57 ± 0.48 g/cc respectively. The CeO2 and SrO2 aerosols are found to be nearly spherical whereas the structural steel aerosols are found to be chain-like structures.

The subchannel code ATHAS-LMR has been improved for analyzing thermal-hydraulics characteristics of flow blockage events in the fuel assembly of a sodium-cooled fast reactor (SFR). Based on the in-depth study of the phenomenon of flow blockage and combined with the characteristics of fast reactor assembly, enhanced models for the wire-wrap spacer and convective heat transfer have been added to account for changes in transverse flow in the presence of blockages. And appropriate transverse mixing and local resistance model are also implemented in ATHAS-LMR-FB. The improved code has been assessed against experimental data obtained at adiabatic and diabatic conditions in SFR assemblies with simulated flow blockages. Good agreement between predictions and experimental data on velocity and temperature distributions downstream of the blockage has been observed. Comparison of predictions of various analytical tools have also been performed. Predictions of the improved ATHAS-LMR code are shown to be compatible with those of other analytical tools. An in-depth analysis of effects of location and size of blockage has been carried out for the fuel assembly of the China Experimental Fast Reactor. Boiling has been predicted to occur for blockage ratios greater than 55.61% at the central location of the fuel assembly. Peak cladding and coolant temperatures are predicted at the axial location 70% of the rod length (just downstream of the maximum local power position of the non-uniform axial power profile).

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.