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Evaluating severe accidents in Generation IV sodium-cooled fast reactors (SFRs) is challenging due to the eutectic reaction between boron carbide (B 4 C) and stainless steel (SS), which results in boron migration and enhanced neutron absorption. A novel quantitative visualization method using radiative heating to observe eutectic behavior and resulting melt structure during boron migration. Experiments, replicating real-scale B 4 C-SS control rods, were conducted in an integrated test facility under core disruptive accidents (CDAs) conditions, introducing a new approach. This study explores the long duration melting (candling phenomenon) of boron within stainless steel, shedding light on relocation behavior not previously studied. We identified two distinct failure mechanisms: the separation of SS from the B 4 C pellet, resulting in the formation of a later melting drop, and the fracture of the B 4 C pellet into multiple pieces, possibly due to thermal stress. The visualization technique and thermal interfacial resistance analysis precisely captured the eutectic temperature, aligning well with prior research. Our study yield insights into eutectic melt relocation behavior by systematically characterizing relocated solidified eutectic melt with a cooling rate of 50 °C/s in stagnant argon environment via scanning electron microscopy-energy dispersive X-ray spectroscopy (SEM-EDS), X-ray diffraction (XRD), and microhardness testing. Our study shows that due to high chromium content, (Cr,Fe) 2 B phases are observed in XRD analysis instead of just Fe 2 B phase. Other phases present in our solidified eutectic melt include γ-Fe (fcc), (Cr,Fe) 23 (B,C) 6 (cubic), and (Cr,Fe) 5 B 3. The atomic composition within these zones was ascertained through SEM-EDS analysis. Our study shows varying hardness values in distinct phases, reflecting differing boron concentrations within micrograph zones. Chromium promotes (Cr,Fe) 2 B and (Cr,Fe) 5 B 3 boride formation, transforming the structure from tetragonal to orthorhombic with increased chromium content, thereby increasing hardness. This validates B 4 C-SS eutectic mixture relocation and diverse boride phase formation under extreme reactor conditions. [ABSTRACT FROM AUTHOR]

• A new solid body formation model in particle method was developed. • The solid bodies in the new model are formed by inter-particle attraction force. • The new model can simulate creation, formation, and motion of solid bodies. • The solid bodies by the new model can diminish and disappear by melting. • The mechanism based on the inter-particle attraction force is common with that in real world. In severe accidents of SFRs, molten core materials can discharge from the core, and the jet can impinge on the lower structure plate. After the jet impingement, fragmented discharged materials can form ring-shape solidification. A fundamental experiment was conducted to simulate the behavior. In order to simulate the behavior of solid body creation and motion, a new solid body formation model by inter-particle attraction force in particle method was developed. The advantage of the new model is that it can simulate creation, formation, and motion of solid bodies without any artificial treatment as solid bodies. The movable solid bodies by the new model have any size, shape, and number, and they are created and grown by solidification, and diminish and disappear by melting. The mechanism based on the inter-particle attraction force is common with that in real world where interatomic attraction force is the cause of solid body formation. [ABSTRACT FROM AUTHOR]

An ingenious prototype Fuel Assembly Inner DUct Structure (FAIDUS) design is conceptualized under the Generation IV safety philosophy to prevent large-scale corium pool formation following total coolant flow blockage accident. The early failure of the duct wall provides a built-in guideway for molten material relocation out of the core region, ensuring inherent safety robustness. Therefore, understanding the duct wall failure mechanism is critical for evaluating accident mitigation potential of the FAIDUS design. In this regard, a transient enthalpy-based multiphase computational model is developed to investigate the heat transfer and fluid flow behavior involved during the duct wall failure in 3-D domain. The numerical model is employed initially to simulate wall failure in the EAGLE ID1 in-pile test. The model results are validated against the in-pile test data. Subsequently, FAIDUS duct wall failure in a prototype fast reactor fuel subassembly (SA) is investigated employing the present model. The study reveals that the large heat flux from the molten pool to the duct is the main reason for the duct wall failure. The duct failure starts with an initial rupture across the corners of the hexagonal-shaped duct and progresses to total failure of the duct wall by the expansion of the rupture over the whole duct wall area with time. The outer hexcan wall sustains only minor thermal damage during the duct wall failure process. The present study explores the associated thermal-hydraulics, molten pool dynamics, crust formation behavior, and event sequences involved in duct wall failure in great detail. The simulation findings suggest that the FAIDUS SA design facilitates premature failure of the slim duct wall prior to the hexcan wall failure. This could potentially enable early discharge of the molten corium away from the core region, limiting the accident progress. • A comprehensive 3-D enthalpy-porosity based multi-phase thermal hydraulic model is developed. • The model is validated against EAGLE ID1 in-pile experimental data. • The rates of damage progression in inner duct and hexcan wall of a prototype SFR fuel SA due to flow blockage are predicted. • Early failure of the thin inner duct and its efficacy in effective relocation of molten fuel out of active core are studied. [ABSTRACT FROM AUTHOR]

• 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]

In this work, a Rotational Fuel-shuffling Breed-and-Burn fast reactor with nitride fuel and sodium coolant (RFBB-NS) was analyzed. The purpose of the study was to design an RFBB-NS, including assembly duct and control rod assemblies, and to show the feasibility. Monte Carlo simulation code Serpent was used for neutronic and fuel burnup analyses and COMSOL Multiphysics was used for heat removal analysis. The results showed that the practical design of RFBB-NS cannot reach criticality due to the higher presence of absorbing materials such as assembly ducts. Therefore, the core was modified to increase the neutron economy by applying a lead–bismuth reflectors. This allowed the reactor to reach criticality and an equilibrium state after fuel shuffling. Afterward, burnup performance analysis showed that the average fuel discharge burnup was ∼ 298 MWd/kgHM and, that the Pu concentration in the discharge fuel was lower than in previous studies. However, the radiation damage to the cladding material was slightly higher than the constraint value in this study. [ABSTRACT FROM AUTHOR]

Hydraulic Suspended Passive Shutdown Subassembly (HS-PSS) is a passive nuclear safety technology employed in Sodium-Cooled Fast Reactors (SFR). During Sodium Fast Reactors (SFR) operation, this component is often subjected to stress cycles caused by flow-induced vibration (FIV), which may lead to material failure and potentially threaten the reactor's safety. Therefore, it is necessary to investigate and evaluate the FIV behavior of the HS-PSS in fast reactors. Different from the conventional used before, HS-PSS is a new type of control rod assembly with no existing studies on FIV. In this study, we established a full-scale experiment for FIV of the HS-PSS in different flow conditions and with various movable component positions (600 mm, 800 mm, 1200 mm). Modal experiments were conducted to determine the natural frequencies. FIV test results show that the highest amplitude occurs when the movable component is 800 mm, with a noticeable acceleration response near the first natural frequency. To explain the experimental observations, numerical simulations have been conducted to analyze the HS-PSS flow field characteristics. The simulations have considered different positions of the movable component at a fixed flow rate of 16.281 m3/h and different flow rates with the movable component positioned at 800 mm. The results indicate significant changes in the flow distribution within the HS-PSS and the hydraulic forces applied to the inlet surface of the movable component when the movable component is set at 800 mm. The experimental and numerical analysis results of this study have provided valuable insights for analyzing and assessing the FIV behavior of the HS-PSS in Sodium-Cooled Fast Reactors. • A test bench was constructed to study the flow-induced vibration of a Hydraulic Suspended Passive Shutdown Subassembly. • Experiments have discovered the location of special movable parts that respond significantly to vibrations. • The flow field simulation results are consistent with the experimental phenomena, explaining the cause of the vibration response. [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]

• A single high-temperature strength equation expressing the mechanical strength in different deformation and rupture modes was derived for 9Cr-ODS TMS cladding tubes. • This equation can predict the rupture life of the cladding tubes under various stresses and temperatures over time. • The reason why the equation can be applied to different deformation and rupture modes is considered to be the effect of the fine-grain matrix of 9Cr-ODS TMS. • The study suggested the thermal activation process is dominant even for the high-temperature deformation modes exceeding yield stress in the 9Cr-ODS TMS with the fine-grained matrix. This paper discusses the applicability of Straalsund et al.'s technique for combining the Larson–Miller parameter (LMP) and life-fraction rule to form a single high-temperature strength equation for 9Cr-oxide-dispersion-strengthened (ODS) tempered martensitic steels (TMS). It uses the extensive dataset on creep rupture, tensile, and temperature-transient-to-burst tests of 9Cr-ODS TMS cladding tubes in the α-phase, α/γ-duplex, γ-phase matrices, which are accumulated by the Japan Atomic Energy Agency so far. The technique is adequately applicable to 9Cr-ODS TMS cladding tubes. A single high-temperature strength equation expressing the mechanical strength in different deformation and rupture modes (creep, tensile, temperature-transient-to-burst) is derived for 9Cr-ODS TMS cladding tubes. This equation can predict the rupture life of the cladding tubes under various stresses and temperatures over time. The applicable range of the high-temperature strength equation is specified in this study and the upper limit temperature for the equation is found to be 1200 °C. At temperatures higher than 1200 °C, the coarsening and aggregation of nanosized oxide particles and the γ to δ phase transformation are reported in previous studies. The high-temperature strength equation can be well applied to the creep and tensile strength in the α-phase matrix, the creep strength in the γ-phase matrix and the temperature-transient-to-burst strength in both phases except for the low equivalent stress (43 MPa) at temperatures exceeding 1050 °C. The mechanism of the notable consistency between creep and tensile strength in the α-phase matrix is discussed by analyzing the high-temperature deformation data in the light of existing deformation models. The study suggested the thermal activation process is dominant even for the high-temperature deformation modes exceeding yield stress in the 9Cr-ODS TMS with the fine-grained matrix. [ABSTRACT FROM AUTHOR]

HT9 cladding was manufactured through successive cold working and heat treatments, and its creep strains were obtained at various stresses and temperatures in the range of 10–120 MPa and 843–953 K, respectively, for up to 20,000 h. Primary and secondary creep regimes were observed, whereas only a few specimens at extreme conditions experienced a tertiary creep regime and creep rupture. The reference creep models, including the theta projection and modified Garofalo equation for HT9 cladding, did not fit the experimentally obtained creep strain curves; the manufactured HT9 cladding showed higher creep resistance than the reference models. Hence, a new correlation was developed by deriving various constants, such as the activation energy, stress exponent, and material constants, from the measured creep strains. The creep correlation successfully predicted the creep behavior of HT9 cladding manufactured. Obtained extensive creep strains over a range of stresses and temperatures, closely approximate actual operating conditions, and developed correlation in this study are valuable in various applications using HT9, providing essential data for analyzing and estimating creep resistances of cladding during the operation of nuclear reactors. [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]

Thermal stratification is an important phenomenon in the process of natural circulation decay heat removal in sodium-cooled fast reactors (SFR). It affects the establishment of the natural circulation and the flow rate of the natural circulation in a station black out (SBO) accident, and therefore affects the safety of the reactor. In safety evaluation, programs are often required to perform analysis for various conditions and long-term transients. However, it is difficult to meet this requirement by adopting a multi-dimensional method or one-dimensional and multi-dimensional coupling method. Therefore, this paper proposes applying a thermal stratification model to the system analysis program THACS to analyze the thermal stratification phenomenon. First, the thermal stratification model is validated with experimental data. The validated model is applied to analyze a station black out accident in the SFR. The lumped parameter model and thermal stratification model are used respectively for the hot pool. By comparing the calculation results of the two models, the effect of thermal stratification of the hot pool on key parameters such as the natural circulation flow rate of the primary circuit, natural circulation flow rate in the inter-wrapper region, average core outlet temperature, maximum core fluid temperature, and the maximum temperature of cladding was analyzed, and the effect of thermal stratification on natural circulation decay heat removal of the SFR was evaluated. This work can provide a reference for the safety analysis of the reactor and contribute to developing a more conservative safety analysis and evaluation strategy. [ABSTRACT FROM AUTHOR]

• Flow regime in debris bed formation behavior with mixed-size particles is studied. • Good predictability of base model using area mean diameter is confirmed. • Past non-spherical particle extension scheme cannot provide satisfactory results. • New extension scheme for mixed-size particles is suggested and validated. • Different component inertia may cause separation/stratification in reactor accident. During a core disruptive accident of sodium-cooled fast reactors, the fragmented debris released from the reactor core will start to sediment and form debris beds within the lower plenum region of the reactor vessel. To ascertain the characteristics of flow regimes during this debris bed formation process, in our earlier publications based on a great number of experiments conducted using mono-sized spherical and non-spherical solid particles, four typical flow regimes (i.e. the particle-suspension regime, the pool-convection dominant regime, the transitional regime and the particle-inertia dominant regime) were identified. In addition, a flow-regime model, including the base model for predictions of spherical particles and the extension scheme for non-spherical particles, has been successfully developed. In this study, motivated to further understand this behavior at a condition of more realistic particle mixture, a new series of simulated experiments using mixed-size particles is conducted under the gravity-driven release from a nozzle into a static two-dimensional water pool. Through detailed analyses, it is found that the base model, by employing the area mean diameter, can predict the overall regime transition under the condition of mixed-size particles reasonably well. Even if a quantity of effective sphericity is introduced, the previous extension scheme for non-spherical particles lacks the capability to provide satisfactory predictions. To improve the prediction using the form of extension scheme, a new correction scheme based on the volume-mean diameter and the degree of convergence in particle-size distribution is suggested and validated. The conducted analyses in this study also indicate that due to the difference in component inertia, the potential separation and stratification of solid particles are likely to exist during an actual debris bed sedimentation process in a core disruptive accident. [ABSTRACT FROM AUTHOR]

• A robust multi-phase thermal hydraulic model for molten fuel relocation is developed. • Movement of molten fuel in a sodium cooled fast reactor subassembly is analysed. • Penetration length of melt front in sub-channels of fuel subassembly is determined. • Amount of fuel relocated out of core before melt front freezing is quantified. • Design measures to enhance fuel relocation to the core catcher are proposed. A transient, multi-phase enthalpy based computational model is developed to analyse the streaming and freezing characteristics of the molten fuel in a blocked fuel subassembly (SA) during Total Flow Blockage (TFB) at the foot of the SA. During TFB accident progression, a molten material pool is formed following the meltdown of fuel pins. The molten fuel has the potential to relocate axially through the available subchannels in the downward direction. The mathematical model comprises of two modules; a Fluid Dynamics module and a Heat Transfer module. The model adopts both implicit and explicit type of finite difference techniques employing a variable grid system for boundary tracking. The mathematical model has been validated against benchmark experimental data and SIMMER III code prediction reported in open literature. The developed model is employed to simulate the relocation of molten fuel in a prototype fast reactor fuel SA under accident condition. The simulation results indicate that the molten fuel is unable to penetrate the axial blankets in fuel SA without freezing and forming a blockage mainly owing to the rapid freezing of melt. Further, a parametric analysis is carried out to ascertain the influence of critical parameters on the relocation and freezing characteristics. It is observed that the design of the fuel SA has a potential effect on the streaming behaviour and enhancement in fuel relocation is substantially achieved by dwindling the fuel pin size, especially axial length and diameter of the pin, thus preventing the accident progression to the neighbouring SAs. [ABSTRACT FROM AUTHOR]

• Super-COPD code is being developed for design study and safety evaluation of SFR. • Plant dynamics analysis was performed for FFTF Loss of Flow Without Scram test #13. • Multi-channel in core and bowing reactivity models enhanced prediction accuracy. • Reactivity and thermal-hydraulics models of Super-COPD for LOFWOS were validated. The Japan Atomic Energy Agency (JAEA) has been developing a plant dynamics analysis code named Super-COPD for design studies and safety analyses of sodium-cooled fast reactors. Since there are a few validation studies of a thermal–hydraulic model in a plant, including a point kinetics model considering reactivity feedback for anticipated transient without scram (ATWS) events classified into the beyond design basis events (BDBEs), further validation studies under BDBE conditions are needed to improve the prediction accuracy of the model. Therefore, JAEA has participated in the international benchmark analysis of loss of flow without scram (LOFWOS) test #13 performed at the Fast Flux Test Facility (FFTF) as one of passive safety demonstration tests hosted by the International Atomic Energy Agency. In the first step, the transient calculation with major reactivity feedback mechanisms and imposing measured power history was conducted with Super-COPD. The results were compared with the measured data. Several challenges were identified about the thermal–hydraulic model in the core and heat transport systems and the reactivity model in the point kinetics model. Consequently, the transient calculation was performed with modified models considering the radial heat transfer, flow redistribution, and interwrapper flow in the core. The time-dependent change and the second peak of the outlet temperature in the specified assembly were predicted accurately because the core thermal–hydraulics was reproduced well. Furthermore, the assembly bowing reactivity model was developed to improve the net reactivity, thereby resulting in a good agreement with the measured data on the reactor power, temperature, and flow rate in the primary loop. Thus, this paper proves the validation of Super-COPD for the LOFWOS event in the BDBE. [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]

• Behavior of in-place cooling of residual core fuel in SFRs was analyzed by SIMMER-III. • Phenomena in the in-place cooling were identified by PIRT. • Out-of-pile experiment simulating the in-place cooling was conducted. • Out-of-pile experiment was analyzed by SIMMER-III. • New model of gas mixing was developed and improved the numerical analysis. In severe accidents of SFRs, the cooling of the residual core materials, which is called "in-place cooling", is one of the important factors for In-Vessel Retention (IVR). For the evaluation, behavior of the in-place cooling was analyzed by the SIMMER-III code. In order to understand the in-place cooling, method of Phenomena Identification and Ranking Table (PIRT) was applied. Based on the result, an out-of-pile experiment which focused on the extracted factors was conducted. In the experiment, continuous oscillation of sodium level was observed by sodium vaporization and condensation. Analysis for the out-of-pile experiment was conducted by SIMMER-III, but the results were different between the experiment and the analysis. By investigation of the analysis result, it was revealed that the difference was due to occupation of non-condensable gas. Therefore, an analysis model of inter-cell gas mixing was newly developed, and the agreement was significantly improved by the new model. [ABSTRACT FROM AUTHOR]

• Validation of THACS and OpenMC for SFR has been enhanced by simulating FFTF LOFWST Test #13. • A new modeling method multi-parallel channel plus local subchannel model is proposed for reactor core. • Important local phenomenon like inter wrapper flow, thermal stratification is analyzed by the specific model. In order to support collaborative efforts within international partnerships on the development of sodium-cooled fast reactor (SFR), the International Atomic Energy Agency (IAEA) proposes a coordinated research project (CRP) about the unprotected loss-of-flow-without-scram test (LOSWOS) performed at Fast Flux Test Facility (FFTF) in 2018, which could validate simulation tool and numerical models for SFR. As one of the participants, the Nuclear THermal-hydraulic Laboratory at Xi'an Jiaotong University (NuTHeL) participates with the monte carol code OpenMC and system thermal hydraulic code THACS. This paperdeal with the standalone neutronic modeling by OpenMC and thermal hydraulic modelling by THACS. For neutronic part, the axial homogenous model was modeled to analyze the reactivity feedback coefficients and steady radial power distribution. For thermal hydraulic part, the system model, including 2D CFD outlet plenum model, subchannel model and inter-wrapper flow model, was modeled to analyze both system response and local phenomena during the LOSWOS. In general, both neutronic results and thermal hydraulic results agreed well with other participants or experiments, the reactivity feedback of GEM could provide major negative feedback, which could be used as a novel passive shutdown device. However, the remaining discrepancies were the underestimated peak temperature of PIOTA at the begin of transients, it may be caused by unknow heat transfer coefficients around hexagon tube, which has no correlation and should be further investigated by experiments. In addition, when considering IWF model, the maximum temperature of the outlet for the active fuel region was 10 K larger than the subassemblie' outlet in the PIOTA, this phenomenon should be carefully treated in loss of flow accidents of 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]

The strong coupling heat transfer between subassemblies and inter-wrapper gaps is closely related to the safety of the sodium-cooled fast reactor (SFR). Therefore, it is very important to carry out a detailed thermal hydraulic analysis of the core. In this paper, the inter-wrapper channel is modeled using the subchannel analysis method. A core thermal-hydraulic analysis code SACOS-Na (Subchannel Analysis Code Of Safety for SFR) is developed by using SIMPLE algorithm. For the subassembly module, a total of three experiments, including ORNL 19-pin experiment, TOSHIBA 37-pin experiment and KNS 37-pin experiment, are adopted to demonstrate the feasibility of the mathematical and physical models and numerical algorithms of the SACOS-Na as well as the rationality of the selected auxiliary models. For the inter-wrapper gap module, WEISS 2-subassembly experiment and CCTL-CFR 3-subassembly experiment are used to verify not only the simulation capability for backflow and strong inter-subassembly heat transfer, but also the heat structure model and the subchannel modeling approach on inter-wrapper gaps. In addition, the prediction accuracy of SACOS-Na has reached the advanced international level by comparing with the worldwide well-known core thermal hydraulic analysis codes. • The subchannel modeling method is adopted to build the mesh for inter-wrapper channels. • The SIMPLE algorithm is used to solve the problems of the reverse flow and inter-subassembly heat transfer occurred in SFRs. • The universal core thermal-hydraulic analysis code SACOS-Na for SFRs is developed. • The accuracy and reliability of the SACOS-Na are discussed from all aspects. [ABSTRACT FROM AUTHOR]

• Severe accident code, AZORES, is coupled with plant dynamics code for SFRs, ADYTUM. • The coupling is performed through LOHS and LORL simulations. • Accident progression between coupled and single AZORES simulations differs significantly. Accident scenarios for the safety evaluation of sodium-cooled fast reactors (SFRs) are categorized into two groups: Anticipated Transient Without Scram (ATWS) and Loss-of-Heat Removal System (LOHRS). The latter is subdivided into two scenarios—Loss-of-Heat Sink (LOHS) and Loss-of-Reactor Level (LORL)—depending on the initiating event. With the development of SFRs, research on ATWS has been paid much attention to due to its rapid accident progression, which could lead to significant radioactive release. Conversely, the significance of LOHRS has increased in recent decades due to lessons learned from accidents at the Fukushima Daiichi Nuclear Power Plant of Tokyo Electric Power Company. However, compared with ATWS, finding and evaluation tools for LOHRS have been limited. Hence, code development is initiated to enhance the severe accident analysis code AZORES, which the Nuclear Regulation Authority developed for the safety evaluation of SFR accidents. This enhancement is achieved by coupling AZORES with plant dynamics code ADYTUM in order to deal with more detailed thermal hydraulics behavior in reactor cooling system. The paper describes the coupling methodology, focusing on the thermal–hydraulic model as the first step of code development and its application. The coupling has been confirmed to be well achieved through a preliminary analysis of LOHS, comparing the results of coupled and single AZORES. The comparison of the analysis results between LOHS and LORL by the AZORES coupled with ADYTUM has been conducted to identify the key phenomena in accident progression. The comparison shows that the coolant leak area and the pressure drop at the piping of the reactor cooling system affect accident progression in LOHS. Consequently, the coupled AZORES is capable of predicting in-vessel thermal–hydraulic behavior in LOHRS successfully, reflecting accident conditions and plant design appropriately. [ABSTRACT FROM AUTHOR]

• Verification and validation for the accuracy of FASYS is performed. • Basic requirements of FASYS and PIRT in CEFR are established. • Bottom-up, top-down and uncertainty assessment are conducted. • Results indicate FASYS can simulate the accident processes of fast reactor. The fast reactor system analysis code FASYS is developed by CIAE for transient accident analysis in CEFR. By referring to the application scope of FASYS, the basic requirements of FASYS and PIRT in CEFR are established. A comprehensive verification and validation for the accuracy of FASYS is performed, which consist of bottom-up assessment, top-down assessment and uncertainty assessment. The bottom-up assessment applies the static analysis and dynamic analysis to verify the basic function of FASYS. The separate effect tests and integral effect tests are employed in the top-down assessment to validate the availability to simulate model and system. The calculated results confirms that the basic function and model in FASYS are reasonable and credible. The uncertainty assessment results show that FASYS can still calculate IET under the reasonable uncertainty of input parameters, indicating that the FASYS can simulate and analyze the transient accident processes of sodium-cooled fast reactor. [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]

• 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 • It shows almost linear propagation in the uncertainty quantification of criticality during the process neutronics calculation in k inf and k eff. • The most important contributors is the inelastic scattering reaction of U-238. • The relative uncertainties of reactivity are bigger than that of criticality in neutronics calculation. • The relative uncertainties of parameters in the MOX-3600 core are always bigger than the ones in the MET-1000 core. Abstract To quantify the uncertainties in the neutronics calculations of sodium-cooled fast reactor (SFR), a series of sensitivity and uncertainty (S&U) analysis were performed on two typical SFRs, which were a large-size MOX-fueled core and a medium-size metallic fueled core. The analysis was done from the pin-cell level to full core level to investigate the uncertainty propagation in the two-step neutronics calculation scheme. The responses analyzed contained the k inf , homogenized cross-sections, sodium void reactivity (SVR), Doppler constant (K D) and control rod worth (CRW) at the sub-assembly level, and the k eff and power distribution at the full core level. The direct numerical perturbation method was used to get the relative sensitivity coefficients and the 'Sandwich Rule' was used to get the relative uncertainties. The results showed that the relative uncertainties propagated linearly from pin-cell to full core for the k inf and k eff. Relative uncertainties of reactivity like SVR were much bigger than the ones of k inf and k eff. However, the relative uncertainty of power distribution was smaller than other responses. Differences were observed in the S&U analysis between the MOX fueled and metallic fueled SFRs. Softer spectrum of the large-size MOX fueled core caused bigger uncertainty compared to the metallic fueled core. [ABSTRACT FROM AUTHOR]

Highlights • New flexible model for a wide range of Sodium cooled Fast Reactors inside CLASS. • Simplifications options lead to bias lower than 1% on actinides composition. • Precision of the meta-model better than 4% on major actinides in spent fuel. • Breeding ratios of designs covered go from 0.88 (burner) to 1.41 (breeder). • Biggest impact on breeding from blankets, major impact from isotopic composition. Abstract Since Sodium cooled Fast Reactors are present in many scenarios and strategies for the future of nuclear energy while not having a specific design established yet, we created a new fast and flexible model for the dynamic fuel cycle simulation tools CLASS. It includes a depletion meta-model and a fuel loading method based on artificial neural networks. It is able to represent a wide range of Sodium cooled Fast Reactor designs using oxide fuels and a wide range of fuel management strategies within fuel cycle simulation tools. A comprehensive analysis of simplification options has been made in order to choose the right level of complexity for the reference full core depletion calculations performed with the MURE code used for the training of the meta-model. The process from these reference calculations to the final meta-model is explained and a specific focus is given to the operations going from detailed full core depletion results to global results suitable for neural networks training. Details on the creation process for neural networks based predictors, one for each average cross-section, and their training on full core depletion calculations are given as well as the implementation within the CLASS code. The irradiation meta-model achieves good precision on all major and minor actinides present in spent fuel. The designs and loaded fuel covered by the model allow significant burner to strong breeder strategies. A sensitivity analysis shows that the number of fertile blankets is the primary contributor for breeding capabilities, but effects of isotopic composition are also significant. A test scenario illustrates the model capacity to simulate burner and breeder designs. [ABSTRACT FROM AUTHOR]

Abstract Research and development of the physical models and codes of safety analyses have been promoted for loop-type Sodium-cooled Fast Reactors (SFRs). A severe accident code, AZORES has been developing for the evaluation of molten corium behavior and integrity of reactor and containment vessels. AZORES enables the simulation of plant equipment behavior under severe accident conditions. In recent years, much attention regarding the mitigation of severe accident progression has been attracted to In-Vessel Retention (IVR) of SFRs, which has relatively less existing experimental and analysis results than that of the Light Water Reactor (LWR). This paper describes the status of development for main physical models and preliminary verification for IVR simulation mainly with AZORES. For a verification process, the simulation has been conducted for representative Post-Accident Material Relocation (PAMR)/Post-Accident Heat Removal (PAHR) phases subsequent to an Anticipated Transients Without Scram (ATWS) to evaluate the impact of initial conditions on severe accident progression. From the study, it was found that the mass of molten corium and capability of decay heat removal were key parameters for the attainment of IVR. Furthermore, this paper describes the results of Computational Fluid Dynamics (CFD) simulations which have been conducted for molten corium relocation in a control rod assembly and for corium coolability with sodium coolant on the core catcher for the purpose of enhancing the related AZORES physical models. Highlights • Modelling of core degradation for a loop type SFR. • Simulation of post-accident phase related to IVR concept. • Capability of severe accident simulation code. [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 • Solidification effects of Bi–Tin melt jets into water coolant were investigated. • Melt-jet modes were organized in a two-dimensional mode diagram. • Melt-jet modes were separated in the jet-breakup and solidification regime. • A scaling analysis using modified Stephan number and Weber number was proposed. • Meltjet was found to completely fragment during CDAs in SFRs. Abstract We present a scaling analysis approach that estimates the melt-jet behavior during the core disruptive accidents (CDAs) in sodium-cooled fast reactors (SFRs). For the scaling analysis, we injected a simulant metal into water coolant and observed the solidification effects on a melt jet. Melt-jet behavior occurred in the water coolant, and the debris settled in the test section. From the experimental results, we categorized the melt-jet behaviors into four modes. We then constructed a scaling analysis based on the Weber number and the modified Stefan number, and mapped the solidification effects on the melt jet in a two-dimensional mode diagram. Using the scaling analysis, we estimated the melt-jet mode in the previous experiments conducted under the large-scaled prototypic conditions, and that under the postulated condition during CDAs in SFRs. [ABSTRACT FROM AUTHOR]

Highlights • The sensitivity of k eff of parts of existing evaluated experiments is analyzed. • Different methods to distinguish the similar experiments are compared. • Similarity characteristics of sodium-cooled fast reactor are discussed. Abstract We conducted a validation of the newly developed SARAX code for the China Fast Reactor (CFR). Unlike the usual way to validate code using dedicated experimental data, we proposed and implemented a theoretical way using the existing experimental data. This paper introduced a method to select the suitable existing experiments, which are "similar" to the CFR's criticality characteristics. We based our method on the sensitivity and uncertainty (S&U) analysis and the similarity analysis. We analyzed different experiments and discussed different ways to distinguish these experiments. Our results showed that a large-size mixed-oxide-fueled experimental core was necessary to represent the critical characteristics of CFR for the code validation. The method using both sensitivity and covariance was better able to distinguish the difference between the objective core and the other cores. [ABSTRACT FROM AUTHOR]

Highlights • Experiments of electromagnetic pump for liquid sodium were carried out. • A new model to predict developed pressure was proposed based on theoretical analyses. • Optimized design was presented based on the capacity analyses and test results. Abstract The electromagnetic pump (EMP) was a critical device and widely applied in Sodium-cooled Fast Reactors (SFR) to deliver liquid sodium. Annular linear induction pump (ALIP) showed relatively higher efficiency than other pumps. Equivalent circuit method was commonly adopted to design the ALIP. Hence, a theoretical model between flow rate and developed pressure of EMP was proposed after detailed deduction based on equivalent circuit theory. To verify the theoretical model, some tests of ALIP were conducted to investigate the characteristic of volume flow rate and developed pressure. Besides, sodium temperature limitation test was also accomplished to test the heat dissipation ability of winding coils through natural air-cooling, results showed that the limit was about 270 °C. Performance analyses was carried out through proposed model which has been verified by experimental results. It turned out that power frequency has positive effect on the pump, whereas sodium temperature showed converse results. Furthermore, the developed pressure increased fast first and then turn to decrease gradually as the annular flow channel width increases. Finally an optimized design was put forward after combining theoretical analyses and experimental research. Compared to the previous ALIP, the volume flow rate, developed pressure and efficiency increased from 9.07 m3·h−1, 386.96 kPa and 28.6% to 10.0 m3·h−1, 574.55 kPa and 48.11% respectively. [ABSTRACT FROM AUTHOR]

Highlights • Typical flow regimes are reproduced under boiling despite local differences. • Similar effect of particle properties under boiling is confirmed. • Boiling can affect regime boundary and characteristics due to vapor bubbles generated. • Non-monotonous effect of water depth under boiling environment is newly recognized. • Validity of using gas-injection to simulate boiling is justified. Abstract Motivated to further understand the effect of sodium boiling on the debris bed formation behavior that might be encountered during a core disruptive accident of sodium-cooled fast reactors, in this work a series of new experiments has been performed under the bottom-heated boiling condition. It is found that the four kinds of flow regimes (namely the particle-suspension regime, the pool-convection dominant regime, the transitional regime and the particle-inertia dominant regime), as observed from previous non-bubbling experiments, can be generally reproduced under current boiling condition, despite the existence of some local differences. Although a similar influence of particle properties (size, density and shape) on the regime transition is observable, the regime boundary is confirmed to be changed due to the steam bubbles generated from boiling. Even for the experimental cases without regime transition, the generated bubbling is verified to have an evident impact on the bed characteristic quantities during the debris bed formation process. As for the experimental parameter of water depth, different from previous non-boiling experiments, in this work for a given heating power, a non-monotonous effect is found to exist, due to the much diminished bubbling rate at rather higher water depths. To be comparable with previous experiments using gas-injection, based on energy conservation a quantity of effective bubbling rate is suggested. By using this quantity, the validity of gas-injection to simulate sodium boiling is justified, thereby providing us enhanced confidence for future studies over a much larger range of gas velocities. [ABSTRACT FROM AUTHOR]

Abstract The main vessel plays an important role in containing the entire primary sodium for pool-type sodium-cooled fast reactors (SFRs). The Reactor Vessel Cooling System (RVCS) has great effect on cooling the main vessel. However, little attention has been given to the study on transient characteristics of RVCS in the previous SFR research. Thus, a home-made one-dimensional (1-D) code named Reactor Vessel Cooling system Analysis Code for Sodium-cooled fast reactor (VECAS) is proposed to evaluate the thermal-hydraulic characteristics for SFR. The detailed models of the developed VECAS are presented in this paper. Moreover, the developed models have been validated against an experimental study. Numerical data of the main vessel cooling circuit are compared with the measurements of the Demonstration Fast Breeder Reactor (DFBR). The simulation results are in good agreement with the experimental data. Furthermore, the validated VECAS is coupled with the Transient Thermal-Hydraulic Analysis Code for Sodium-cooled fast reactors (THACS). The transient characteristics of RVCS in China Experimental sodium-cooled Fast Reactor (CEFR) are simulated by the coupled code. Steady analysis shows that the main vessel is cooled effectively. The peak temperature appears at the top of the main vessel lower than the permissible upper temperature limit. During the transient analysis, VECAS has predicted a reverse flow in RVCS, which contributes to the core cooling. Furthermore, sensitivity analysis of the main parameter has also been performed. Therefore, it can be concluded that coupled VECAS has the ability to evaluate the thermal-hydraulic characteristics as well as the decay heat removal capacity of RVCS. The coupled code could provide references and technical supports for the design and optimization of the pool-type sodium-cooled fast reactor. Highlights • A one-dimensional (1-D) code named VECAS is proposed to evaluate the thermal-hydraulic characteristics for SFR. • The code has been validated against the main vessel cooling circuit of DFBR. • VECAS is coupled with THACS to calculate the transient characteristics of RVCS in CEFR. • VECAS has the ability to evaluate the thermal-hydraulic characteristics and the decay heat removal capacity of RVCS. [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]

The capability of transient analysis has been implemented in the SARAX code system recently. Alternatively, two calculation schemes using the point-kinetics model and 3-D transport-based spatially-dependent model have been developed. Based on these, two transient benchmarks of sodium fast reactor (SFR) suggested by the Working Party on Reactor and System (WPRS) were calculated, which include the large size core loaded with oxide fuel (MOX-3600) and the medium size core loaded with metallic fuel (MET-1000). Two typical states including the state of begin-of-cycle (BOC) and state of end-of-cycle (EOC) were considered. The unprotected loss of flow (ULOF) transient, unprotected transient over power (UTOP) transient, and unprotected control rod withdrawal (UCRW) were calculated. For the large size MOX-fueled SFR, the ULOF transient is the most severe accident, which leads to the core damage in a very short time. At the EOC state, the thermal margin is smaller because of the change of fuel compositions due to the depletion. Besides, the comparisons show that it is very important to use the spatially-dependent model in predicting the power level in the UCRW transient, where the point-kinetics model significantly underestimates the power level by 7–17%. [ABSTRACT FROM AUTHOR]

To understand the pool sloshing behavior with unmelted core materials during a core disruptive accident of sodium-cooled fast reactors, in our earlier publication a series of simulated experiments was performed using different solid particles. It was recognized that due to the different interaction mechanisms between solid particles and the gas bubble injected, three kinds of flow regimes, termed respectively as the bubble-impulsion dominant regime, the transitional regime and the bed-inertia dominant regime, could be identified. In this work, a regime map is developed to describe the regime transition by considering the competitive role between the bubble impulsion and bed inertia during the sloshing process. It is found that a respectable agreement between experiments and the regime-map predictions could be achieved given current range of conditions including much difference in particle size, particle bed height, particle density and the gas-injection pressure. [ABSTRACT FROM AUTHOR]

Abstract Modelling of sodium-evaporation and formation of sodium-oxide aerosols from a sodium-pool fire is of fundamental importance for the assessing of the radiological consequences in Sodium-cooled Fast Reactors severe accidents. This paper summarizes the derivation of a simple model to estimate the amount and size of particles being generated from Na-pool fires and its performance assessment, once implemented in an integral severe accident tool (ASTEC-Na), against available large-scale separate effect tests. The model has been transposed in analytical correlations which implementation in lumped-parameter severe accident codes is straightforward. According to the comparisons to data set, the correlations do not adversely impact the code estimates with respect to other more empirical alternative approaches and, in addition, the correlations remove any need of user-defined ad-hoc parameters in the input deck concerning Na-based particles behaviour, as other alternates do. Regarding code behaviour, the model predictions yield the same order of magnitude both in terms of suspended aerosol concentration and diameter as data and capture the reliable measured data trends. Highlights • Development of particle generation model for sodium-oxides aerosol formation. • Encapsulation of a zero-D model to correlations compatible to lumped parameter codes. • In-containment particulate source term prediction. • Assessment of the model performance against open large scale experiments. • Implementation of derived correlations to a lumped parameter code (ASTEC-Na). [ABSTRACT FROM AUTHOR]

In sodium-cooled fast reactors, liquid sodium leakage from piping may lead to fire accident. In the case that the liquid sodium is discharged as a liquid jet, a number of droplets are produced during liquid jet impingement on the structures; the surfaces of splashed droplets serve as a main reaction field of sodium combustion. In the present work, a liquid jet was emanated vertically downward from a circular nozzle hole onto a horizontal disk to measure the total amount and the maximum size of splashed droplets. It was found that the splashing rate was negligibly small when the liquid jet impinged as a continuous jet whilst a significant amount of liquid was splashed when the liquid jet impinged as a broken jet. Thus, a prediction method was first developed for the impact frequency of the primary droplets produced due to the jet breakup. It was then shown that a phenomenological model using the impact frequency and the impact Weber number as the important variables can predict the splashing rate well. It was also indicated that the size of the maximum splashed droplets was fairly proportional to the size of primary droplets. [ABSTRACT FROM AUTHOR]

In modern sodium-cooled fast reactor (SFR) designs, paralleling steam generators (SGs) are used to improve the economy. However, large leakage sodium-water reaction (SWR) models were developed only for designs using one SG. It is necessary to derive a pressure wave propagation model with parallel channels. Thus, a detailed pressure wave propagation model based on conversion equations was derived and verified with experimental data in this study. Using the developed model, a large leakage SWR accident in the SFR with four parallel SGs was simulated. The accident process was analyzed, and the design was verified to accommodate a large leakage SWR. Sensitivity analysis of critical parameters on large leakages was simulated and investigated. The SG pipe area, rupture disk bursting pressure, failure of the rupture disk, and release pipe area were found to have significant influences on the maximum pressure. Thus, the findings of this study are valuable for SG designs and protection from large leakage SWR accidents. • A parallel channel pressure wave propagation model is derived. • The dynamic model is verified with experimental data. • The accident process and secondary loop integrity are investigated. • The parameters to indicate the large leakage SWR accident are proposed. • Sensitivity analysis on the large SWR is performed. [ABSTRACT FROM AUTHOR]

Studies on local fuel-coolant interaction in a molten pool are important for the improved evaluation of severe accidents that might be encountered for sodium-cooled fast reactors. To clarify the characteristics of this interaction, several years ago a series of simulated experiments was conducted at the Japan Atomic Energy Agency (JAEA) by delivering a given quantity of water into a molten pool formed with a low-melting-point alloy. In this work, motivated by acquiring more reliable database and knowledge, an improved experimental system has been developed recently at the Sun Yat-sen University. Experimental results from various conditions including much difference in water volume, melt temperature, water subcooling, initial water-lump shape along with melt depth, are discussed. It is found that the water volume, melt temperature, initial water-lump shape and melt depth are observable to have remarkable impact on the pressure-buildup characteristics while the role of water subcooling is less prominent. The observed consistent influence of water volume, melt temperature and water subcooling, as compared to previous studies at JAEA, indicates that despite the existence of some uncertainties, the statistical results recognized at JAEA might be overall valid. Owing to much enriched evidence, the present analyses suggest that the most probable reason leading to the limited pressurization as water volume increases for a given melt and water temperature, should be primarily due to the isolation effect of vapor bubbles at the melt-water interface. [ABSTRACT FROM AUTHOR]

The evaluation of thermal-hydraulics of core under decay heat removal conditions is essential to the safety evaluation of a sodium-cooled fast reactor. The inter-wrapper flow (IWF) has an influence on the thermal-hydraulics of core. However, the study of the flow and heat transfer of IWF was limited. In this paper, a 2D layered IWF model was developed, and some tests were simulated to validate the IWF model, including heat removal tests SHRT-17 and SHRT-45R conducted in the experimental fast reactor EBR-II and a natural circulation test performed during the PHENIX end-of-life experiment. In order to simulate all the components of the primary system, the IWF model is coupled with the Transient Thermal-Hydraulic Analysis Code for Sodium-cooled fast reactors (THACS). In the simulation without IWF model, the predicted peak temperature of the instrumented subassembly XX10 in EBR-II is about 150 K lower than test data, and the predicted average outlet temperature of reactor core in PHENIX is about 20 K higher than test data. While the predictions of THACS with IWF model agree well with the test data. The results show that the IWF can be accurately simulated by the IWF model, and the IWF model improves the accuracy of the simulations of the reactor core. Further, some sensitivity analyses were conducted to provide better reference for the study of sodium-cooled fast reactors. [ABSTRACT FROM AUTHOR]

The FASt TEst Reactor (FASTER) plant is a sodium-cooled fast spectrum test reactor design that was developed as part of an advanced demonstration and test reactor study, and provides high levels of fast and thermal neutron flux for scientific research and development. The 120 MWe FASTER plant has been designed with extended testing capabilities in mind, while keeping the design as simple as possible in order to make it efficient and cost attractive. The main function of the reactor is to provide neutrons for irradiation testing utilizing existing proven technologies to ensure a high technology readiness level. A preliminary safety analysis was performed of the FASTER design using the fast reactor safety analysis code SAS4A/SASSYS-1 to assess the reactor's safety performance during beyond design basis accidents. Three types of transient scenarios were analyzed for the FASTER design: unprotected transient overpower, unprotected loss of heat sink, and unprotected station blackout. These transients were simulated at beginning of equilibrium cycle and end of equilibrium cycle conditions. For each transient, failures of both the primary and secondary reactor scram systems were assumed. Because of the favorable reactivity feedbacks, relatively low peak fuel, cladding, and coolant temperatures were obtained during the transients and the reactor vessel temperature remained well below the limit for ensuring structural stability. [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]

A calculation model was constructed to systematically study the effects of environmental conditions (i.e. Cr concentration in sodium, test temperature, axial temperature gradient of fuel pin, and sodium flow velocity) on Cr dissolution behavior. Chromium dissolution was largely influenced by small changes in Cr concentration (i.e. chemical potential of Cr) in liquid sodium in the model calculation. Chromium concentration in sodium coolant, therefore, should be recognized as a critical parameter for the prediction and management of Cr dissolution behavior in the sodium-cooled fast reactor (SFR) core. Because the fuel column length showed no impact on dissolution behavior in the model calculation, no significant downstream effects possibly take place in the SFR fuel cladding tube due to the much shorter length compared with sodium loops in the SFR plant and the large axial temperature gradient. The calculated profile of Cr concentration along the wall-thickness direction was consistent with that measured in BOR-60 irradiation test where Cr concentration in inlet sodium bulk flow was set at 0.07 wt ppm in the calculation. [ABSTRACT FROM AUTHOR]

Thermal-hydraulic analysis was performed with a 7-pin fuel bundle by using CFD. The ordinary wire-wrapped spacers are wound on the fuel pins in the same direction. Counter flow is predicted to occur in all sub-channels due to this pattern. This flow was confirmed in this analysis work. A new type of arrangement for wire-wrapped spacers, called the U-pattern, is presented to provide favorable flow for coolant mixing. In this pattern, 7 pins are designated in a group, the center pin has no wrapping spacer, and the pins surrounding the center pin have alternate winding directions of the spacer. Superior mixing effect and the uniform flow were confirmed in this pattern. This pattern has about 30 °C cooler than that of the ordinary wire-wrapped fuel bundle pattern in the maximum temperature region. Pressure drop of the U-pattern was reduced about 10% compared with the ordinary pattern. The new pattern of the fuel pins enhances the heat transfer and reduces pressure drop simultaneously. [ABSTRACT FROM AUTHOR]

Studies on the pool sloshing behavior are important for the improved evaluation of energetic potential of a large whole-core-scale molten fuel pool that might be formed during a Core Disruptive Accident (CDA) of Sodium-cooled Fast Reactors (SFR). Motivated to understand the characteristics of this behavior, in this study a series of simulated experiments was conducted by injecting nitrogen gas into a Two-Dimensional (2D) rectangular water pool through a nozzle positioned at the center of pool bottom. To achieve a comprehensive understanding, experimental parameters, including nitrogen gas pressure (∼4.2 bar), initial water depth (∼60 cm), gas injection duration (0.06–0.1 s) along with the nozzle size (10–50 mm), were varied. Through detailed analyses, it is found that under current range of conditions, all the experimental parameters employed are confirmable to have remarkable impact on the sloshing characteristics (e.g. maximum elevation of water level at the pool center and peripheries). The performed analyses also suggest that possibly due to a diminished residence time of the injected-gas in the pool, a limited sloshing intensity is observable as the gas-injection pressure increases. Evidence and fundamental data from our work will be utilized for the empirical-model development as well as the analyses and verifications of future SFR severe accident analysis codes in China. [ABSTRACT FROM AUTHOR]