Accidents In Sodium-cooled Fast Reactors Research Articles

Numerical investigation of SUPERFACT-1 experiment by a MOX fuel pin thermo-mechanical model

Preparations are currently underway to develop ISAA-Na code, a sodium-cooled fast reactors (SFRs) specific version of an integrated severe accident analysis code. This version aims to evaluate both protected and unprotected accidents in SFRs. Given the crucial role of fuel pin thermo-mechanical behavior in SFR safety analysis, an advanced model for MOX(Mixed OXide) fuel pin thermo-mechanical performance under operation conditions has been developed and validated. This validation involved simulating the SUPERFACT-1 fast reactor irradiation experiment using ISAA-Na. The resulting calculation data were compared with experimental measurements and the outcomes of other codes dedicated to fuel pin performance analysis (TRANSURANUS, MACROS, andGERMINAL) to assess the fuel pin thermo-mechanical model. The overall code assessment reveals that the model achieves satisfactory performance in simulating fuel pin behavior, with predicted errors falling within an acceptable range.

Development of a simplified one-dimensional CDA bubble model

Sodium-cooled fast reactors (SFRs) have high safety with an extremely low probability of a core disruptive accident (CDA). However, from a defense-in-depth perspective, the CDA sequence is still worth studying. Severe accidents in SFRs, such as unprotected loss of flow, might lead to a CDA during which fuel and some fission products could be instantly released from a large CDA bubble through potential leak paths in the top shield structure. Therefore, a reasonable prediction of dynamic behavior of a large-scale bubble within the sodium pool is vital for accurately evaluating the migration of source terms. In this study, we propose a simplified one-dimensional CDA bubble model that can handle heat and mass transfer in two-phase multicomponent materials in different computational domains during the rising of bubbles through the sodium pool toward the cover-gas region. In this model, an entrainment model based on the Rayleigh–Taylor instability and Kelvin–Helmholtz instability was used to explain the coolant entrainment through the gas/liquid boundary and jet fragmentation. The model also addresses the mitigation effect of non-condensable gases on the condensation of fuel, steel and sodium vapor at bubble interface. We evaluated the model using a past experiment on the expansion of a two-phase, large bubble with high pressure in a stagnant liquid pool conducted by Purdue University in the late 1970 s using a 1/7-scale model of Clinch River Breeder Reactor. Good agreement with the experimental data demonstrates that the developed model can reasonably represent the essential characteristics of dynamic behavior of a large, high-pressure bubble with heat and mass transfer in two-phase multicomponent materials. This is valuable for evaluating the migration of the source terms, which will be carried out in future studies.

Thinning behavior of solid boron carbide immersed in molten stainless steel for core disruptive accident of sodium-cooled fast reactor

ABSTRACT Boron carbide (B4C)–stainless steel (SS) eutectic reaction behavior is one of the most important issues in the core disruptive accidents (CDAs) of sodium-cooled fast reactors (SFRs). In this study, the immersion experiments using B4C pellets with molten SS were conducted to evaluate the CDA sequences such as contact event of solid B4C with degraded core materials including SS at very high temperature. The immersion experiment aims at understanding the kinetic behavior of solid B4C–liquid SS reaction based on the reduced thickness of B4C pellet after the experiment in the temperature ranges from 1763 to 1943 K, which is higher than the temperature of solid B4C–solid SS reaction. Based on the kinetic consideration of the reaction rate constants for solid B4C–liquid SS reaction, it was found that similar temperature dependency was identified between solid B4C–liquid SS and solid B4C–solid SS. Besides, the reaction rate constants of solid B4C–liquid SS were smaller than those of solid B4C–solid SS extrapolated in higher temperature region by two or more orders of magnitude due to two different evaluation method for B4C side/SS side. It was confirmed that this difference was reasonable through the consideration of previous reaction tests in solid–solid contact for B4C side/SS side.

Numerical study of debris bed formation behavior for severe accident in sodium-cooled fast reactor by using least square MPS-DEM method

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.

Experiment and new analysis model simulating in-place cooling of a degraded core in severe accidents of sodium-cooled fast reactors

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.

Spatio-temporal evolution of sodium aerosol and temperature distribution during sodium pool fire in a confined environment

In case of a Core Disruptive Accident in sodium-cooled fast reactor, sodium will be ejected from the reactor vessel into the Reactor Containment Building (RCB) and results in sodium fire. Liquid sodium combustion generates aerosol, which gets dispersed and remains suspended in RCB based on geometry of the building and degree of turbulence. The Environment Source Term (EST) primarily depends on the dynamics of sodium aerosols inside the containment. Hence, it is essential to carry out a safety analysis of SFR using computational tools and models shall be validated with experimental data, particularly aerosol dynamics inside the containment along with thermal hydraulics. Besides, all codes assume uniform aerosol concentration in large containment buildings and the evolution of suspended concentration was predicted. However, in actual case a non-uniform concentration profile exists, which leads to stratification of aerosol concentration inside the containment. Towards assessment of aerosol dynamics in realistic scenario, spatial–temporal characteristics of sodium aerosol and temperature distribution were studied by conducting sodium pool fire experiments in a large containment.The sodium fire experiment was conducted by using ∼2.0 kg of sodium in a pool area of ∼0.25 m2. The measured maximum pool and gas temperatures are 685 °C and 126 °C respectively, and the temperature gradient at the wall boundary layer was 7.5 °C/cm. The estimated average sodium burning rate and aerosol release rate are 17.3 kg/m2hr and 0.24 g/s, respectively. The spatial variation of aerosol characteristics was measured for both horizontal and vertical planes. The measured maximum aerosol concentration in the horizontal plane ranges from 2.76 to 5.32, 2.16–3.30 and 2.09–4.21 g/m3 for the bottom, middle and top elevations respectively. For the vertical plane, aerosol concentration varied from 2.16 to 3.65 g/m3 for central positions and 2.53–5.32 g/m3 for off-center positions. The trend of average suspended aerosol concentration at each elevation showed that the concentration was higher at the top elevation than at the middle and bottom for up to 15 min i.e., during aerosol generation period and after that concentration was found to be higher at the bottom elevation for the next 25 min. Based on the measurements, it is observed that the aerosol distribution is not well mixed within the experimental period. A significant difference in aerosol mass concentration was observed between each elevation, between estimation and observations based on a zero-dimensional model. The evolutions of aerosol median size by theoretical simulation capture the overall trend reasonably.

Characteristics and Mechanisms of Debris Bed Formation Behavior in Severe Accidents of Sodium-Cooled Fast Reactors: Experimental and Modeling Studies

A Sodium-cooled Fast Reactor (SFR) is one of the optimized candidates in Generation IV nuclear reactor systems, but safety is an essential issue for SFR development and application. Most knowledge was accumulated through SFR safety investigations, especially for Core Disruptive Accidents (CDAs). During the CDA of SFRs, the molten materials in the core region are likely to discharge into subcooled sodium and form a debris bed on the lower region of the reactor vessel. Noticing that elaboration on the characteristics and mechanisms of Debris Bed Formation (DBF) behavior should be essential for the subsequent analysis of debris bed coolability and accident progression through various experimental and modeling studies, much knowledge was obtained during the past decades. Motivated to promote future investigations on CDAs of SFRs, the previous experiments and modeling studies on DBF behavior are systematically reviewed and discussed in this paper. The experimental results showed that the flow-regime and accumulated-bed characteristics during DBF were influenced by varying parameters and realistic conditions. Through the modeling studies, several empirical models were proposed for predicting the flow regime and accumulated-bed characteristics in DBF. In addition, to promote further development of research, the future prospects concerning DBF behavior are also described.

Thermal Characteristic Analysis of Sodium in Diluted Oxygen via Thermogravimetric Approach

As the main reactor type of the fourth-generation nuclear power systems, sodium-cooled fast reactors are now designed and built worldwide. A sodium pool cooling circulation process is indispensable in a sodium-cooled fast reactor. However, the sodium pool fire design is the basis of accidents in sodium-cooled fast reactors. The fire hazard caused by the sodium–oxygen reaction and fast reactor safety have attracted extensive attention. Dry powder is widely used as an effective fire-extinguishing agent to control sodium fire. The sodium will burn in an oxygen-depleted atmosphere when using dry powder to cover fire. In this study, the change law of thermogravimetry of melted sodium is studied by thermogravimetric analysis (TGA) and the apparent activation energy (Ea) is obtained, which has a linear relationship with the oxygen concentration. The results can provide a reference for improving the engineering design standards of sodium fire suppression systems and can also be incorporated into simulation software to improve the accuracy of fire suppression simulations.

Numerical simulation of jet breakup phenomenon during severe accident of sodium-cooled fast reactor using MPS method

During a core disruptive accident of Sodium-cooled Fast Reactors (SFR), discharged corium may contact with coolant in form of jet, which can be an imminent threat to the integrity of reactor vessel if the jet does not sufficiently break up and cooldown. It is significant to ascertain the morphologies, penetration depth and breakup mechanism of discharged corium jet to evaluate its potential impact on reactor vessel. In this study, numerical simulations are carried out to investigate the characteristics of jet breakup. An advanced moving particle semi-implicit method (MPS), the least square MPS (LSMPS) is applied to study the two-dimensional hydrodynamic corium jet breakup in sodium. Firstly, the applicability and reliability of the method are validated by the comparison between the simulation results and reference experiments. Then, the method is applied to simulate molten corium jet breakup in sodium. It is found that the breakup mechanism is related to Rayleigh–Taylor instability (RT), Kelvin–Helmholtz instability (KH) and Plateau–Rayleigh instability (PR). Under the influence of RT and KH, eddies are formed around the jet and thus the downward velocity of the jet vibrates in a relatively regular pattern. The local velocity difference of the jet narrows the jet and the jet eventually breaks up under the influence of PR. The parameters including diameter and initial velocity of the jet were scrutinized for their influences on jet penetration depth and jet breakup length, which can be useful for the improved design of the SFR reactor vessel.

Preliminary Analysis of Core Disruptive Accident in Sodium-Cooled Fast Reactor Using Eutectic Reaction Model of Boron-Carbide Control-Rod Material

It is necessary to simulate a eutectic melting reaction and relocation behavior of boron carbide (B4C) as a control rod material and stainless steel (SS) during a core disruptive accident (CDA) in an advanced large-scale sodium-cooled fast reactor (SFR) designed in Japan. A physical model simulating the eutectic reaction and relocation of the eutectic melt was developed to incorporate into the fast reactor severe accident analysis code SIMMER-IV for the CDA numerical analysis of SFRs. This study applied the SIMMER-IV code with the newly developed model to the CDA analysis of the SFR. This analysis indicated that the SIMMER-IV code using the eutectic reaction model has successfully simulated the eutectic reaction and the upward motion of the eutectic melt in the molten core pool as well as the reactivity transient behavior caused by the molten core material relocation.

Experiment and Analysis for Development of Evaluation Method for Cooling of Residual Core Materials in Core Disruptive Accidents of Sodium-cooled Fast Reactors

Experiment and Analysis for Development of Evaluation Method for Cooling of Residual Core Materials in Core Disruptive Accidents of Sodium-cooled Fast Reactors

Experimental Study on Reaction Behavior between Control Rod Material and Molten Stainless Steel for Core Disruptive Accidents of Sodium-cooled Fast Reactors

Experimental Study on Reaction Behavior between Control Rod Material and Molten Stainless Steel for Core Disruptive Accidents of Sodium-cooled Fast Reactors

Numerical evaluation of hypothetical core disruptive accident in full-scale model of sodium-cooled fast reactor

A hypothetical core destructive accident (HCDA) has received widespread attention as one of the most serious accidents in sodium-cooled fast reactors. This study combined recent advantages in numerical methods to realize realistic modeling of the complex fluid–structure interactions during HCDAs in a full-scale sodium-cooled fast reactor. The multi-material arbitrary Lagrangian–Eulerian method is used to describe the fluid–structure interactions inside the container. Both the structural deformations and plug rises occurring during HCDAs are evaluated. Two levels of expansion energy are considered with two different reactor models. The simulation results show that the container remains intact during an accident with small deformations. The plug on the top of the container rises to an acceptable level after the sealing between the it and its support is destroyed. The methodology established in this study provides a reliable approach for evaluating the safety feature of a container design.

Severe Accidents in Sodium-Cooled Fast Reactors: Computational Fluid Dynamics Sensitivity Studies on the Thermal Ablation of Core Catcher

Abstract The present work aims at testing the computational fluid dynamics capabilities to simulate erosion of materials which interact with a corium mass. The main foreseen application is the design of external/internal core catchers or in-vessel retentions devices used to mitigate severe accidents for sodium-cooled fast reactors. 2D axisymmetric simulations of a corium jet impinging a sacrificial solid material show evidence of a pool-effect, previously observed in experiments, which contributes to limit the ablation process. Complementary sensitivity calculations assess the influence of jet diameter, temperature, and velocity.

구리밀봉 증기발생기 설계코드의 개발과 이를 이용한 TRU 함유 노심계통의 증기발생기 설계

KAERI has been conducting R&D on sodium-cooled fast reactor(SFR) as a next-generation reactor, and a copper bonded steam generator(CBSG) has been proposed to minimize the occurrence of sodium-water reaction(SWR) accident in SFRs. In this paper, a development progress of a CBSG sizing analyzer(CBSGSA) code is described. A sizing and rating analysis of CBSG has been performed to set thermal-hydraulic conditions applicable to the steam generator of the TRU core system.

A Study on Sodium–Concrete Reaction in Presence of Internal Heating

Abstract The sodium–concrete reaction (SCR) is an important phenomenon during severe accidents in sodium-cooled fast reactors (SFRs), as it generates large volumes of hydrogen and aerosols in the containment vessel along with structural concrete ablation. In this study, the chemical reaction beneath the internal heater (800 °C) was investigated in SCR experiments with internal heating. The experiments simulate the effects of obstacles and heating on the SCR. Especially, we focused on the concrete ablation phenomenon because the hydrogen generation is sourced from the moisture in the concrete. The effects of internal heating on the self-termination mechanism are also discussed. The internal heater on the concrete hindered the transport of sodium (Na) into the concrete. Therefore, the reaction between Na and the concrete began at the periphery of the internal heater, where the concrete ablation depth was larger than under the internal heater. The high Na pool temperature (800 °C) largely increased the Na aerosol-release rate, which was explained by Na evaporation and formed a porous reaction-product layer. The Si mass balance and image mapping by an electron-probe micro-analyzer yielded consistent porosities in the reaction-product layer (0.54–0.59). The porous reaction products suppressed the amount of Na transported into the reaction front. Regardless of the internal heater placement, the Na concentration around the reaction front was limited to around 30 wt %. The Na concentration condition was dominantly responsible for the self-termination of the internally heated SCR.

Boiling incipience of liquid sodium flow in an annular channel based on wavelet analysis

During hypothetical accidents of sodium-cooled fast reactors, such as unprotected loss of flow, transient power surge accidents, unprotected thermal trap accidents, etc., sodium boiling may occur. A series of boiling incipient phenomena for liquid sodium flow in an annular channel were investigated experimentally. The boiling incipience signal in different experimental conditions were analyzed through the Discrete Wavelet Transform (DWT). The various differential pressure signals in the process of boiling incipience were collected by a differential pressure transducer installed on the effective heating section. The experiment was performed in the heat flux range of 128–846 kW/m2, mass flow rate range of 7.2–122.0 kg/h, system pressure range of 0.85–28.79 kPa and inlet subcooling range of 63.1–287 °C. Since these correlations of predicting the boiling incipience for water in different channels are not suitable for the boiling incipience of liquid sodium, a semi-empirical correlation for the prediction of boiling incipience is obtained in the present study. The prediction error of the new semi-empirical correlation for experimental data is basically within ± 30% in this paper. This work provides a meaningful liquid sodium boiling incipience prediction method.

Validation for Eutectic Melting Model of Boron-Carbide Control-Rod Material in Core Disruptive Accidents of Sodium-Cooled Fast Reactors

Validation for Eutectic Melting Model of Boron-Carbide Control-Rod Material in Core Disruptive Accidents of Sodium-Cooled Fast Reactors

Numerical investigation of flow blockage accident in SFR fuel assembly

Sodium-cooled fast reactor (SFR) is one of the most promising candidates to meet the Generation IV International Forum (GIF) declare. Due to the good thermo-physical properties of sodium, SFR has drawn a lot of attentions and there are considerable design experiences related to SFR. In the design of SFR, the pitch to diameter ratio (P/D) is kept small. Debris may accumulate in the flow channels and result in the flow blockage accident due to the corrosion of structural material. If single or several sub-channels are partially or totally occupied by debris, flow area will be suddenly reduced and a recirculation region can be expected downstream the blockage. Coolant temperature in this region can be greatly increased which may threaten the integrity of fuel assembly. In this paper, RANS method is employed to evaluate the impaired heat transfer process governed by blockages formed in a sodium-cooled 61-pin wire-wrapped rod bundle. The simulation results were firstly validated against the experimental data obtained from the literature. A good agreement demonstrated the capability of the employed approach to evaluate turbulent properties of sodium flow as well as the impaired heat transfer characteristics caused by blockages. With the validated method, the influence from blockages are systematically evaluated in which blockage size and its location are varied. Simulation results clearly indicate that a sharp increase of cladding and coolant temperature near the blockage can be expected even if the blockage area is not quite large. The local effect from blockage is also identified in the simulation. The variance of cladding and coolant temperature tends to vanish downstream the blockage and it is not feasible to detect the existence of blockage based on cladding temperature near the outlet even if the blockage area is sufficiently large.

Investigation of flow-regime characteristics in a sloshing pool with mixed-size solid particles

Abstract To ascertain the characteristics of pool sloshing behavior that might be encountered during a core disruptive accident of sodium-cooled fast reactors, in our earlier work several series of experiments were conducted under various scenarios including the condition with mono-sized solid particles. It is found that under the particle-bed condition, three typical flow regimes (namely the bubble-impulsion dominant regime, the transitional regime and the bed-inertia dominant regime) could be identified and a flow-regime model (base model) has been even successfully established to estimate the regime transition. In this study, aimed to further understand this behavior at more realistic particle-bed conditions, a series of simulated experiments is newly carried out using mixed-size particles. Through analyses, it is verified that for present scenario, by applying the area mean diameter, our previously-developed base model can provide the most appropriate predictive results among the various effective diameters. To predict the regime transition with a form of extension scheme, a correction factor which is based on the volume-mean diameter and the degree of convergence in particle-size distribution is suggested and validated. The conducted analyses in this work also indicate that under certain conditions, the potential separation between different particle components might exist during the sloshing process.