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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]

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]

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]

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]

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]

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]

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 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]

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]

Sodium cooled fast reactor (SFR) exhibit some unique thermo-hydraulic characteristics in comparison to thermal reactors, considering that SFR have a higher core power density, special configuration of assemblies, wire-wrapped spacers, special thermo-physical properties of sodium and so on. Some design limits related to fuel, cladding and coolant temperature must be met in both the normal and transient condition. Thus, it is of great importance to obtain accurate thermal-hydraulic performance for the design and safety assessment of the SFR reactor. Therefore, a subchannel analysis code SUBAC specially for SFR wire-wrapped assemblies is developed in the present study. Sensitivity analysis of different model combinations is carried out by using ORNL 19-pin benchmark test. The calculated results show that the normalized temperature at the exit of the rod bundle calculated by the Cheng-Todreas correlation combined with Cheng-Tak turbulent mixing model is consistent with the experimental data. Subsequently, the 19-pin and 37-pin bundle benchmark problems are calculated to validate the accuracy and applicability of the code. All the calculated results agree well with the experimental data, indicating that the code is competent for steady-state calculations. Finally, SUBAC code is used for performing transient analyses of EBR-II SHRT-17 XX09 subassembly. A good agreement between calculated and experimental values of the core top sodium temperature is achieved, but slightly underestimated. In conclusion, the SUBAS program is capable of being used for subchannel analysis of SFR and some suggestions for future research are given. [ABSTRACT FROM AUTHOR]

Studies on debris bed formation behavior are of crucial importance for the improved evaluation of Core Disruptive Accidents (CDA) that could occur for Sodium-cooled Fast Reactors (SFR). In this work, to clarify the mechanisms underlying this behavior, a series of experiments was performed by discharging solid particles into two-dimensional rectangular water pools. To obtain a comprehensive understanding, various experimental parameters, including particle size (0.256∼8 mm), particle density (glass, alumina, zirconia, steel and lead), particle shape (spherical and irregularly-shaped), water depth (0–60 cm), particle release pipe diameter (10–30 mm), particle release height (110–130 cm) as well as the gap thickness of water tank (30–60 mm), were varied. It is found that due to the different interaction mechanisms between solid particles and water pool, four kinds of regimes, termed respectively as the particle-suspension regime, the pool-convection dominant regime, the transitional regime and the particle-inertial dominant regime, are identified. The performed analyses in this work also suggest that under present experimental conditions, the particle size, particle density, particle shape, particle release pipe diameter and water depth are observable to have remarkable impact on the above regimes, while the role of particle release height and gap thickness of water tank seems to be less prominent. Knowledge and data from this work might be utilized for the improved design of core catcher as well as analyses and verifications of SFR severe accident analysis codes in China in the future. [ABSTRACT FROM AUTHOR]

Experiments simulating overheating conditions of fast reactor severe accidents have been previously carried out with irradiated fuels. For the present study, the chemical forms of the fission products (FPs) included in the irradiated fuels were evaluated by thermochemical equilibrium calculations. At temperatures of 2773 K and 2973 K, the most stable forms of Cs, I, Te, Sb, Pd and Ag are gaseous compounds, and their fractions are about 100% in an oxygen partial pressure range between 10 −5 and 10 3 Pa. Cs and Sb detected in the thermal gradient tube (TGT) in the experiments can take gaseous chemical forms of elemental Cs, CsI, Cs 2 MoO 4 , CsO and elemental Sb, SbO, SbTe, respectively. The chemical forms of Cs and Mo are sensitive to the oxygen partial pressure. By comparing experimental results and the estimations, it is seen CsI thermochemically behaves in a manner that traps it in the TGT, while elemental Cs trends to move as fine particles. The moving behavior of the gaseous FPs will obey not only thermochemical principles, but also those of particle dynamics. [ABSTRACT FROM AUTHOR]

Much of the basic technology for the Sodium-cooled fast Reactor (SFR) has been established through long term development experience with former fast reactor programs, and is being confirmed by the Phénix end-of-life tests in France, the restart of Monju in Japan, the lifetime extension of BN-600 in Russia, and the startup of the China Experimental Fast Reactor in China. Planned startup in 2014 for new SFRs: BN-800 in Russia and PFBR in India, will further enhance the confirmation of the SFR basic technology. Nowadays, the SFR development has advanced to aiming at establishment of the Generation-IV system which is dedicated to sustainable energy generation and actinide management, and several advanced SFR concepts are under development such as PRISM, JSFR, ASTRID, PGSFR, BN-1200, and CFR-600. Generation-IV International Forum is an international collaboration framework where various R&D activities are progressing on design of system and component, safety and operation, advanced fuel, and actinide cycle for the Generation-IV SFR development, and will play a beneficial role of promoting them thorough providing an opportunity to share the past experience and the latest data of design and R&D among countries developing SFR. [ABSTRACT FROM AUTHOR]

Abstract: Reasonable mathematical and physical models as well as auxiliary models have been established to develop a subchannel analysis code for one fuel assembly of the Sodium-cooled Fast Reactor (SFR). The conduction model of mixed fuel UO2–PuO2 was adopted in the sub-channel analysis. The comparison of the flow velocity distribution in the fuel assembly was performed between the Chiu-Rohsenow-Todreas (CRT) and Novendstern models. Heat transfer correlations for liquid metals were compared with each other and one was selected as the optimization correlation. The validation of the code was performed with Oak Ridge National Laboratory (ORNL) 19 pin tests. The temperature profiles at the end of the heated length for low and high power cases were compared between experimental results and other codes. And then, based on the subchannel code, thermal-hydraulic characteristics of the Chinese Experimental Fast Reactor (CEFR) were investigated. Axial and radial coolant temperature profiles for different subchannels were presented. In addition, the mass flow rate with mixing effects were carefully studied. The effect of the wire was investigated and the optimization ratio of the pitch to diameter was provided according to current simulated conditions. [Copyright &y& Elsevier]

Abstract: The Sodium-Cooled Heterogeneous Innovative Burner Reactor (SCHIBR) model created at the University of South Carolina uses heterogeneous minor actinide targets. To improve minor actinide transmutation, a hybrid fuel management scheme is utilized involving initially moderated assemblies on the core periphery followed by a second period of irradiation in a fast flux with the moderating rods removed. A multi-tiered recycling strategy was developed to increase plutonium utilization in the SCHIBR model through the recycle of the driver fuel. An equilibrium fuel cycle was evaluated with the computer code ERANOS to determine the improvements in fuel utilization, reduction in high level waste, and safety of the SCHIBR design. Fuel depletion studies were conducted to determine the composition of input and output streams in order to develop reactor recipes for use in the fuel cycle simulation code, VISION. The once-through SCHIBR model reduces the radiotoxicity of high level waste by 66% of the once-through LWR model after 300 years in storage. The multi-tiered recycling strategy offers improvements over the previous once-through SCHIBR model by reducing the radiotoxicity by 86% after 300 years in storage. [Copyright &y& Elsevier]

Abstract: This paper presents a concept of the dual tier system consisting of the existing light water reactor (LWR) plants and sodium-cooled fast reactor (SFR) for transuranics (TRU) burning for the purpose of downsizing the required SFR. In this system, Pu is combusted by the LWR at first and then the remaining Np, Am, and Pu are destructed by the SFR. The iteration number of Pu combustion by the LWR is chosen to be twice owing to the sodium void reactivity limitation of $6. As a result of combustion calculation, the twice Pu burning of LWR lessens the TRU amount by 27% and changes the composition significantly. Moderator pins of zirconium hydride are deployed to the SFR fuel subassembly so as to enhance TRU burning and reduce the sodium void reactivity. The nuclear calculation found that the core characteristics become similar to the conventional SFR due to the moderator: the sodium void reactivity remains still $4 and the Doppler coefficient becomes −6 × 10−3 Tdk/dT. This study concludes that this dual tier strategy can downsize the required SFR to approximately 40% of the single tier system of SFR with TRU conversion ratio of 0.6. [Copyright &y& Elsevier]

Abstract: 4S (Super-Safe, Small and Simple) is a small sized sodium-cooled fast reactor being developed for the electricity supply in remote areas, high-temperature steam supply more than 400 °C, seawater desalination, and hydrogen production. The system design of power output of 10 MWe (30 MWt) has been completed. The main feature is that it does not have to be refueled for a long period (i.e. 30 years for 10 MWe version), and enable the reactor closure sealed during plant operation. Furthermore, the small size of the reactor makes the reactor building suitable for below grade installing. These two features can provide resolutions for the issues relevant to safety, security, and safeguard, which become much more serious matter internationally these days. 4S is a pool-type reactor which contains the whole primary cooling system in a vessel. For the purpose of reducing the maintenance requirements with the reactor, (1) reflectors to compensate for fuel burn-up instead of control rods, (2) electromagnetic pump (EMP) which has no rotating parts, and (3) residual heat removal system by natural circulation and natural air draft are adopted. Therefore, exchange of the reactor components is not required during plant operation, in addition to no needs for refueling. Toshiba has initiated the U.S. Nuclear Regulatory Commission (NRC) pre-application review of 10 MWe version for the purpose of applying for design approval (DA). A series of public meetings with NRC has been held four times, and five technical reports have been submitted to NRC in preparation for DA application. Topics discussed in these meetings included, plant design, metallic fuel, safety design philosophy, safety analysis, measures against severe accident, phenomena identification and ranking table (PIRT), etc. Some useful comments and questions on the issues regarding the specific feature of 4S as well as sodium-cooled fast reactor were raised by NRC at the public meetings. Among them, those items which are applicable to general sodium-cooled fast reactors, e.g. principal design criteria, guideline for safety analysis, validation and verification for safety analysis code, quality requirements, severe accident, and emergency planning are presented in this paper. [Copyright &y& Elsevier]

Abstract: A conceptual sodium-cooled fast reactor (KALIMER-600) has been developed at KAERI. In this work, the simplified performance analysis code for the KALIMER-600 was developed and the power maneuvering characteristics without control rod movement was evaluated. We developed the analysis code by modifying a commercial Modular Modeling System (MMS) code. We attached the library for the properties of sodium as an additional user library and modified some component modules such as the fuel and pipehx with specific features of KALIMER-600. Then, the power maneuvering capability without control rod movement was evaluated when the BOP (balance of plant) power was changed. We called the power maneuvering capability with inherent reactivity feedback mechanism without control rod movement as self-control. We, in turn, evaluated the self-control capability under varied and/or fixed the flow rate of the coolant systems of the plant. We concluded the KALIMER-600 had a good self-control capability and some data obtained from this study would be used in developing control rod programming and flow rate programming of plant during power maneuvering. [Copyright &y& Elsevier]

A sodium-cooled fast reactor (SFR) is one of Generation IV nuclear systems. The once-through steam generator is important for heat transfer and acts as a barrier to separate the sodium and water. If the heat transfer tube breaks, a sodium–water reaction (SWR) occurs, and the integrity of the secondary loop may be destroyed. Thus, a dynamic model to analyze the SWR must be developed, especially for a large-leak SWR. A large-leak SWR analysis model was developed for an SFR, consisting of four modules: water/steam leakage rate, hydrogen bubble growth, pressure wave propagation, and SWR protection system action. A large-leak SWR accident was simulated using this model, and the process was investigated. Sensitivity analysis was performed for the critical parameters, and the parameters with a key influence on the large-leak SWR were determined. The results are significant for the design of the steam generator and the SWR protection system. • A large-leak SWR analysis method was developed. • The large-leak SWR of the SFR was simulated and analyzed. • Sensitivity analysis was performed to investigate the critical parameters on SWR. • The results are of significance to design of steam generator of the SFR. [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]

Sodium-cooled Fast Reactor (SFR) is an optimized candidate of the next-generation nuclear reactor systems, while safety is a critical issue during its Research and Development (R&D) process. Therefore, SFR safety analysis, especially for severe accidents such as the Core Disruptive Accident (CDA), has received great attention over the past decades. It was noticed that as the CDA of SFR proceeds, the molten-pool sloshing motion, which can greatly influence the recriticality, is possible to occur. Due to its great importance for the improved assessment of CDAs of SFR, a large amount of knowledge concerning this issue has been accumulated in the past decades through various experimental and modeling studies. In this paper, the past experimental and modeling investigations on the molten-pool sloshing motion under various conditions are systematically reviewed and discussed with the aim to provide a valuable reference for the future CDA study and safety assessment of SFRs. Furthermore, to promote deeper research and exploration for this issue, some future remarks are also pointed out. Image 1 • Past experimental & modeling studies reviewed for molten-pool sloshing motion in CDA. • Studies on sloshing motion compared and discussed under various different conditions. • Empirical modeling approach and ideas introduced and discussed for sloshing motion. • Future prospect provided for further researches on molten-pool sloshing motion. [ABSTRACT FROM AUTHOR]

To analyze the process of a small-leak sodium–water reaction (SWR) in the secondary loop of the steam generator (SG) in sodium-cooled fast reactors (SFRs), a multi-module program is developed herein. This program is developed by evaluating the models of heat transfer tube self-wastage, target tube wastage, and hydrogen migration and dissolution, which are involved in the small-leak SWR accident caused by a heat transfer tube with leakage in a single-module SG. The developed program could be used to investigate the time of occurrence of the heat transfer tube self-wastage, adjacent tube wastage, and the response time of the hydrogen concentration detectors at different leakage rates. The program can further provide the reasonable positions of hydrogen concentration detectors. The results of this study provide a reference for designing the secondary loops of small-leak accident protection systems for SFRs. Image 1 • A program is developed to simulate phenomena induced by small leak sodium water reaction accident. • Self-wastage and target tube wastage models, hydrogen source and dissolution models are proposed. • Hydrogen concentration detector layout is studied for small leak accident protection system. [ABSTRACT FROM AUTHOR]

The thermal stratification phenomenon plays a crucial role in the safety of various nuclear systems, including the Gen-III + Light Water Reactors (LWR) and the Gen-IV reactors. The phenomenon is of particular importance for the pool-type Sodium-cooled Fast Reactors (SFRs) because it may cause neutronic and thermal-hydraulic instabilities in the reactor core, or lead to damages of both the reactor vessel and in-vessel components due to the growth of thermal fatigue cracking. More significantly, thermal stratification could impede the establishment of the natural circulation during accidental scenarios and introduce uncertainties to the core safety of SFRs. Efforts for modeling of the thermal stratification in SFRs have been made for decades to prevent or mitigate the damage caused by the phenomenon. This paper gives a review of the advances that have been made in recent 10 years on the computational modeling methods for thermal stratification phenomenon in SFRs. These methods can be generally drawn into two categories. The first one is the system-level methods which provide fast-running but approximate calculations, and the second one is the CFD methods which provide high-resolution calculations at high computational expense. After introducing the efforts that have been made to improve the one-dimensional (1-D) models, the paper envisioned the possible research directions that could be pursued to enhance the modeling of thermal stratification in the near future. • Outstanding experiments and associated analytical findings considering thermal stratification are introduced. • Modeling efforts on the thermal stratification in SFRs made in the last 10 years are reviewed. • Possible research directions to enhance the modeling of thermal stratification in the near future are envisioned. [ABSTRACT FROM AUTHOR]

Investigations of the sloshing motion in molten pool with unmelted/refrozen core materials are crucial to improve the evaluation of severe accidents that are possibly encountered for a sodium-cooled fast reactor. Based on a great number of experiments mainly conducted with spherical particles, in our earlier work, three types of flow regimes (namely the bubble-impulsion dominant regime, the transitional regime and the bed-inertia dominant regime) were recognized and a regime map (original model) has been even established successfully to estimate the transformation of regimes for spherical particles at varied parametric conditions. In this article, motivated by further checking the impact of particle shape on the sloshing motion, a series of new experiments has been conducted using different non-spherical particles. It is recognized that compared to spherical particles, the regime boundary for experiments using non-spherical particles tends to emerge within a relatively bigger particle size. To predict the regime characteristics for non-spherical particles, a correction scheme which consists of modifications of the original model from several aspects is suggested. Through detailed analyses and discussion, we find that with the proposed correction scheme coupled, our developed model can provide satisfactory predictions on the transition of flow regimes for the pool sloshing motion with both non-spherical and spherical solid particles over our present scope of experimental conditions. • New experiments on pool sloshing motion using non-spherical particles performed. • Regime boundary for non-spherical particles found at larger particle size. • Correction scheme successfully proposed to incorporate particle-shape effect. • Good predictions obtained for non-spherical & spherical particles using the scheme. [ABSTRACT FROM AUTHOR]