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Summary: Closed type assembly leads to complicated core thermohydraulics which are significant for design and safety analysis in sodium‐cooled fast reactor (SFR). Most of the major phenomena can be dealt with by general parallel channel network and two‐dimensional model for inter‐wrapper flow (IWF) in system code transient thermal‐hydraulic code for analysis of SFR (THACS). A main deficiency left is the overestimation of inter‐assembly heat transfer effect due to lump temperature message that one‐dimensional assembly model can only provide. A methodology was thus developed to compute wall‐subchannel temperature for simulating energy transportation through duct wall. Intra‐assembly thermohydraulics were classified as two mechanisms and non‐dimensional parameter was introduced for characterizing each of them. Ratio between linear heat flux density through hexagonal wrapper surface and pin surface qw/p was employed to account for radial heat transfer. Flow redistribution effect arouse in mixed and natural convection was concretized by a importance symbol of buoyancy force, Richardson number Ri. Two criteria were also proposed for dividing flow state into three parts and specific model for each section was defined. Two validation cases against 3‐D computational fluid dynamics (CFD) simulation of China experimental fast reactor whole core and test data from plant dynamics test loop with direct heat exchanger experiment were performed with THACS. Noticeable improvement after incorporation of this model demonstrated its value in system analysis. Although the procedure for parameter computations is geometry related, this method can be reproduced to any type of assembly directly. [ABSTRACT FROM AUTHOR]

Summary: To evaluate residual heat removal capacity of next generation China sodium‐cooled fast reactor (SFR) and provide data for code validation, an integral natural circulation experiment was performed on a scaled water platform which had same structure with reactor primary system. A modular code THACS focusing on safety analysis of SFR was validated by this test. Some necessary development and verification for physical property, frictional loss, and heat transfer of water were carried out first. Whole simulation of an unprotected loss‐of‐flow (ULOF) case was divided into two phases and computation lasted until 5000 s of test time during second stage. Agreement in light of system temperature distribution in first stage provided a suitable initial condition for subsequent transient. Satisfied flow rate curve was reproduced with decaying pressure head and valve characteristic as input when pump was stopping. All natural circulation paths were predicted, including original flow in primary loop through main pipe and reversed flow in reactor vessel cooling system (RVCS). Transition to operative mode of direct heat exchanger (DHX) in direct reactor auxiliary cooling system (DRACS) and its cooling power were also forecasted successfully. This work means a significant advance for THACS in the field of engineering application of China SFR. [ABSTRACT FROM AUTHOR]

Understanding the Debris Bed Formation (DBF) behavior is important to the optimized integral evaluation for Core Disruptive Accident (CDA) of Sodium-cooled Fast Reactor (SFR). Aimed at clarifying the characteristics of flow regimes during DBF, extensive experimental works were carried out by releasing solid particles into a water pool. According to the preliminary studies using bottom heating and nitrogen gas injection, it was found that the effect of coolant boiling should be crucial to the flow-regime transition. However, the gas flow rate considered in these studies (0–50 L/min) was not adequate wide to cover the range of vapor flow rate from the possible drastic boiling in actual CDA of SFR. Therefore, in current study, to elucidate the influence of coolant boiling on DBF flow-regime characteristics more comprehensively and provide more extensive data for supporting the predictive model development, a greater number of experiments are conducted with a much-enlarged range of gas flow rates (up to 100 L/min) under varying parametric conditions (including particle size, particle type, water depth and release pipe diameter). Through analyzing the flow-regime transition, an extended functional form for the effect of coolant boiling is successfully proposed. By incorporating this form into our base model, the model predictability can be extended for the DBF at the conditions of coolant boiling induced by accumulated debris in accordance with the verification of the regime map. Present study is believed to be valuable for further developing and verifying the models and codes for SFR safety analysis in China. • Experiments of debris bed formation in CDA of SFR are conducted with gas injection. • Qualitative observations and varying quantitative parametric analyses are performed. • Effect of coolant boiling induced by accumulated debris is studied and confirmed. • An empirical model extension function is developed based on extensive experiments. • The predictability of the extended model is validated through the regime map. [ABSTRACT FROM AUTHOR]

• Effect of interaction mode on local FCI checked experimentally using same couple of melt-coolant materials. • Significant cooling/re-condensation of steam bubbles by surrounding water pool confirmed at MI mode. • Impact of liquid volume delivered, water subcooling and pool depth is totally different within two modes. • Conversion efficiency at MI-mode is significantly lower compared to CI-mode. From the past analyses of Core Disruptive Accidents (CDA) for sodium-cooled fast reactors, it has been recognized that with the assumption of very pessimistic conditions, a large-scale molten-fuel pool, which contains sufficient amount of fuel to achieve the prompt criticality by fuel compaction, might be formed. Due to the potential entrapment of liquid sodium into the fuel pool, local Fuel-Coolant Interaction (FCI) in the pool is considered to be one of the various initial factors resulting in such compactive fluid motions. To ascertain the mechanisms underlying this interaction more reliably, in our earlier publication a number of simulated experiments was carried out by releasing a given amount of liquid water into a molten pool (comprised of a low-melting-point alloy) using the PMCI (Pressurization characteristics in Melt-Coolant Interaction) facility recently developed at the Sun Yat-sen University. In this work, motivated to acquire more enhanced understanding, the effect of interaction mode on local FCIs is further investigated by adding new experimental data within the Melt-Injection (MI) mode. From detailed analyses, it is found that except the existence of a similar impact of melt temperature, much difference in the influence of other experimental parameters such as the liquid volume delivered, water subcooling along with the pool depth is observed. In addition, the performed analyses in this work also confirm that compared to previous Coolant-Injection (CI) mode, the mechanical energy conversion efficiency is much lower within the MI-mode condition. This work provides a large palette of beneficial experimental data and evidence for the better understanding of CDAs and improved verifications of fast reactor safety accident analysis codes in China. [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]

China Fast Reactor (CFR) with 600 MW electric power is under development in China. Lots of efforts are being made to its Research and Development (R&D). The CFR600 steam generator with straight tubes consists of the evaporator and the superheater. Phase change exists in the water/steam side and the thermal–hydraulic behavior is difficult to predict. It is challenging to design such a steam generator. Therefore, a suitable thermal–hydraulic code is necessary to assist its design and analysis. A steady-state thermal–hydraulic model is derived based on the mass, energy and momentum equations. The characteristic tube is divided into four regions: subcooled region, nucleate boiling region, film boiling region and superheated region. Equal node enthalpy increment in a region is applied to enhance the numerical accuracy. The model is verified with the design results of design code SG-33 and CEFR. It is proven that the code developed can be used to design and analyze the thermal–hydraulic behavior. Sensitivity analysis is performed to investigate the influence on the heat transfer from key parameters. A preliminary design of CFR600 steam generator is presented using the developed code and reasonable results are obtained. [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 China Experimental Fast Reactor (CEFR) is a small, sodium-cooled fast reactor with 20 MW(e) of power. Start-up tests of the CEFR were performed from 2010 to 2011. The China Institute of Atomic Energy made some of the neutronics start-up-test data available to the International Atomic Energy Agency (IAEA) as part of an international neutronics benchmarking exercise by distributing the experimental data to interested organizations from the member states of the IAEA. This benchmarking aims to validate and verify the physical models and neutronics simulation codes with the help of the recorded experimental data. The six start-up tests include evaluating criticality, control-rod worth, reactivity effects, and neutron spectral characteristics. As part of this coordinated research, the IAEA performed neutronics calculations using the Monte Carlo codes Serpent 2 and OpenMC, which can minimize modeling assumptions and produce reference solutions for code verification. Both codes model a three-dimensional heterogeneous core with an ENDF/B-VII.1 cross-section library. This study presents the calculation results with a well-estimated criticality and a reasonably good estimation of reactivities. The description and analysis of the core modeling assumptions, challenges in modeling a dense SFR core, results of the first phase of this project, and comparative analysis with measurements are presented. [ABSTRACT FROM AUTHOR]

Hydraulically Suspended Passive Shutdown Subassembly (HS-PSS) of Sodium-cooled Fast Reactor (SFR), which can protect the reactor from beyond design basis accidents, is especially suitable for responding to the Unprotected Loss of Flow accident (ULOF). In this study, the performance of the conceptually designed HS-PSS for China Fast Reactor with 600 MW electric power (CFR600) that is being developed by the China Institute of Atomic Energy (CIAE) was experimentally investigated. Suspension tests, free drop performance tests, and hydraulic characteristic verification of the HS-PSS were carried out on a specially constructed vertical hydraulic test loop. Furthermore, the hydraulic characteristic data such as suspension critical flow rate, drop time, pressure drop, and hydraulic thrust on the Hydraulically Moving Body (HMB) inside the HS-PSS was obtained to verify whether they meet the design requirements. The HMB prototype has a drop time of 10.65s under the coast-down flow rate, and the initial falling critical flow rate is 1.80 kg/s, at 84 °C. Both agree well with the theoretical calculation results and meet the engineering design requirements. • The performance of the HS-PSS prototype of CFR600 was experimentally investigated on the VERCH test facility in water. • The critical flow rate of HMB suspension in the upper working position at 84 °C is 1.8 kg/s. • The drop time of HMB under the coast-down flow at 84 °C is 10.65s, which meets the design requirements. • Experimental validation qualifies the calculation program developed. [ABSTRACT FROM AUTHOR]

• New experiments conducted by delivering water into molten LBE pool. • Similar trend in transient behavior and parametric effect confirmed for two melts. • Despite lower thermal diffusivity, water cooling on surrounding LBE is limited. • Pressurization in LBE experiments is larger due to enhanced melt density. • Compared to static pressure, melt depth is affected more by melt penetration. In a Core Disruptive Accident (CDA) of Sodium-cooled Fast Reactor (SFR), by supposing rather pessimistic condition, a large molten fuel pool would be possibly formed that contains enough fuel with the potential to exceed prompt criticality due to fuel compactive motion. Local fuel-coolant interaction (FCI) in the molten pool is identified as one of the typical inducing factors that may result in such compactive motion. To understand the characteristics of this interaction, owing to the PMCI (Pressurization characteristics in Melt-Coolant Interaction) facility recently established at the Sun Yat-sen University, in our earlier publications two series of simulated experiments, i.e. Coolant-Injection (CI) experiments and Melt-Injection (MI) experiments, have been conducted, in which water and low-melting-point Bi-Sn-In alloy (60% Bi, 20% In and 20% Sn) were used respectively as the simulant materials for sodium and molten fuel. In this study, aimed at checking the generality of the experimental findings as well as achieving further enhanced understanding on this interaction, a large number of experiments are newly performed through delivering water into a simulated molten fuel pool comprised of Lead-Bismuth Eutectic (LBE) alloy (another low-melting-point alloy). Through detailed comparisons (with the CI-mode experiments using Bi-Sn-In alloy), it is found that a similar trend in the transient pressure and temperate history as well as the overall effect of experimental parameters (such as water quantity, melt and water temperature, water-lump shape and melt depth) on the pressurization characteristics can be reproduced for the new LBE experiments. Despite a much lower thermal diffusivity, for the LBE experiments it is confirmed that the most likely reason resulting in the limited pressurization as water volume increases should be also primarily owing to the isolation effect of vapor bubbles generated at the melt-water interface. Possibly due to the enhanced melt density which facilitates the penetration of melt through vapor bubbles, under the same parametric condition the pressurization in LBE experiments is generally much higher than that in Bi-Sn-In experiments. Compared to melt static pressure, the pressurization under different melt depths is supposed to be more influenced by the varied easiness of melt penetration. Current work provides a large amount of favorable experimental database and insight for promoted understanding on actual CDAs and improved verifications of SFR severe accident codes in China. [ABSTRACT FROM AUTHOR]

• Melt-jet behavior studied with quantified particle and bed properties (esp. debris shape). • Melt & water temperature impacts breakup length, fragment shape & bed porosity. • Melt velocity & nozzle size affect little fragment shape and bed porosity. • Evident effect of water depth on fragment shape and bed porosity confirmed. • Diagrams predicting melt-jet mode, debris sphericity and bed porosity suggested. Investigations on melt-jet behavior during Core Disruptive Accidents (CDAs) of Sodium-cooled Fast Reactors (SFR) are crucial to the design of In-Vessel Retention (IVR) strategy. In present work, aimed to achieve a comprehensive understanding on this behavior, a number of simulated experiments are conducted by discharging a molten low-melting-point alloy jet into a water pool. Different experimental parameters such as melt temperature, water temperature, melt penetration velocity, nozzle diameter and water pool depth are taken. Based on visual observation as well as the quantified particle and debris-bed properties measured (e.g. average size and size distribution of fragments, particle sphericity and bed porosity), the effect of different releasing conditions on the melt-jet behavior mode, particle characteristics and debris-bed morphology are clarified. Dominant factors that influence the melt-jet breakup phenomenon are identified over present range of conditions. The performed analyses also suggest that the fragment mass median diameter can be reasonably interpreted by the Kelvin-Helmholtz instability model which integrates solidification effect and the Weber number theory. Besides, map diagrams distinguishing hydrodynamic effect and thermal effect with regard to melt-jet behavior mode, debris bed morphology (porosity) and particle morphology (sphericity) are explored. Insight and experimental data obtained in this work will be utilized for the future improved validations of fast reactor safety analysis codes in China, such as the thermal–hydraulic models for the prediction of fragmentation and debris formation phenomena. [ABSTRACT FROM AUTHOR]

• Effect of all parameters on sloshing confirmed under liquid stratification. • Limited sloshing intensity verified using different liquid or with stratification. • Except injection duration, all other parameters can affect critical gas pressure. • Empirical model for predicting critical gas pressure successfully developed. • Fuel compaction in reactor accident with liquid stratification should be mitigated. Investigations on the sloshing motion of a molten fuel pool are of great significance for the improved assessment of core disruptive accidents that might occur for Sodium-cooled Fast Reactors (SFR). In this work, aimed to acquire further evidence and insight for understanding the mechanisms of this motion at more realistic complicated conditions, a large number of simulated experiments are conducted newly by importing nitrogen gas into two-dimensional liquid pools of different properties (esp. including 70 cases under the condition of liquid stratification). To achieve deeper and more comprehensive understandings, various experimental parameters such as the gas-injection pressure (~4.5 bar), pool depth (~60 cm), liquid density (Idemitsu cleaning oil, purified water), depth ratio of stratified liquids and the injection duration (0.06 ~ 0.1 s), are taken. Through detailed analyses, it is confirmed that all our present experimental parameters employed can have remarkable impact on the sloshing intensity. With increasing the gas-injection pressure, the tendency of limited sloshing intensity as observed from a pure water pool in our earlier work, can be generally reproduced at current conditions (e.g. in a pool with different liquid density or with the stratification of two liquids). Except the parameter of injection duration, the remaining experimental parameters (i.e. the pool depth, liquid density and the depth ratio of stratified liquids) are verified to have an obvious influence in changing the critical gas pressure that is required to achieve the limited intensity of sloshing. Based on the knowledge obtained from experimental analyses, an empirical model is successfully established to estimate the critical gas pressure. The performed analyses in this study also imply that comparing with pure single-phase molten pool, due to the entrainment of molten fuel by the upward gas-induced flow, the degree of fuel compaction should be mitigated under the existence of liquid stratification. Experimental data and knowledge from our work will be employed for the future improved validations of fast reactor safety analysis codes in China. [ABSTRACT FROM AUTHOR]