Reactor structural response to molten fuel-coolant interactions

In order to realistically determine the structural response of a liquid metal fast breeder reactor to a molten fuel-coolant interaction (MFCI), an MFCI region was incorporated into the two-dimensional, hydrodynamic containment code, REXCO-H. In this way, it is possible to account for the two-dimensional hydrodynamic response, as well as for the effect of vessels and plates, upon the expansion process in the MFCI region.The MFCI model has been extended in order to increase the usefulness of the code under a variety of conditions. The sodium equation of state has been improved using basic thermodynamic relations and recent data to incorporate temperature dependent properties. Heat transfer models available to describe the MFCI include not only a quasi-steady-state model, but also a parametric model, including the fuel heat of fusion. Nonhomogenous MFCI regions can be treated by assigning different parameters to each zone within a region, including volume fractions of fuel, sodium, steel, and void, as well as initial fuel and coolant temperatures and fraction of molten fuel.Several cases have been studied in order to delineate the effect of various parameters on the peak pressures generated in the MFCI zones. These include effect of initial fuel and coolant temperatures, void fraction, amount of molten fuel and/or vessel wall compliance. The response of a typical reactor configuration is evaluated for a given set of initial conditions.     

Fragmentation of molten debris during a molten fuel-coolant interaction

The potential for an energetic molten fuel-coolant interaction (MFCI) during a hypothetical core meltdown accident is of concern in nuclear safety analysis. An important aspect of a MFCI is the fine fragmentation and intermixing of molten core debris with the core coolant. The fragmentation characteristics of the molten debris (a mixture of UO₂ and zircaloy cladding) particles produced during a recent high-energy in-pile experiment are analyzed. The experimental results suggest that two mechanisms contributed to the fragmentation of the molten debris in this experiment, in which an MFCI occured. Phenomenological modelling of these two mechanisms and the effects of the governing parameters are presented.     

Molten fuel-coolant interaction during a reactivity initiated accident experiment

The results of a reactivity-initiated accident experiment, designated RIA-ST-4, are discussed and analyzed with regard to molten fuel-coolant interaction (MFCI). In this experiment, extensive amounts of molten UO₂ fuel and zircaloy cladding were produced and fragmented upon mixing with the coolant. Coolant pressurization up to 35 MPa and coolant overheating in excess of 940 K occurred after fuel rod failure. The initial coolant conditions were similar to those in boiling water reactors during a hot startup (that is, coolant pressure of 6.45 MPa, coolant temperature of 538 K, and coolant flow rate of 85 cm³/s). It is concluded that the high coolant pressure recorded in the RIA-ST-4 experiment was caused by an MFCI and was not due to gas release from the test rod at failure, Zr/water reaction, of UO₂ fuel vapor pressure. The high coolant temperature indicated the presence of superheated steam, which may have formed during the expansion of the working fluid back to the initial coolant pressure; yet, the thermal-to-mechanical energy conversion ratio is estimated to be only about 0.3%.     

Assessment of Fuel Coolant Interactions (FCIs) in the FBR Core Disruptive Accident (CDA)

Application of general behavior principles (GBPs) and consideration of relevant contact modes suggest that only incoherent small-scale fuel coolant interactions (FCIs) with negligible damage potential appear possible with the molten oxide fuel-liquid sodium system as the fuel disperses away from the core into a coolable non-critical array.

In contrast to the SPERT-1, BORAX-1 and SL-1 nuclear transients that ultimately led to energetic vapor or steam explosions, the presence of molten fuel and liquid sodium in the FBR core always requires the presence of solid cladding which separates the fuel and coolant and, hence prevents energetic FCIs prior to coolant escape.

Furthermore, unlike the CORECT-II experiments which examined dynamic re-entry of liquid sodium on molten fuel pools that resulted in unstable interfaces leading to significant sodium entrapment and relatively energetic FCIs, the prevailing contact mode in the FBR core disruptive accident (CDA) scenario is displacement of the lighter and less viscous liquid sodium by the heavier and more viscous molten fuel resulting in stable interfaces with no significant sodium entrapment and FCIs. Dynamic re-entry of liquid sodium into the core is not possible with the two-component steel vapor-liquid sodium system, since the interface contact temperature upon steel vapor condensation is well in excess of the sodium boiling temperature. A pressure reduction in the steel vapor region due to condensation is immediately compensated for by an equivalent pressure increase due to sodium evaporation.

Finally, considering that the molten oxide fuel-liquid sodium interface contact temperature is well below the sodium homogeneous nucleation temperature which in turn is well below the fuel melting temperature, not only eliminates the potential for large-scale vapor explosions as molten fuel streams are injected into liquid sodium pools, but also implies that small scale superheat explosions are possible which are consistent with the usually observed incoherent sharp pressurization events (amplitudes up to the order of 10 MPa and duration of the order of 1 ms). These general behavior characteristics are also consistent with complete fuel fragmentation with fragment sizes ranging from 100 to 1,000 μm, and the absence of significant or damaging FCIs.     

Fundamental Experiment on Vapor Explosion with Mass of Grains

With the analysis of hypothetical accident in a nuclear power reactor, a molten fuel and coolant interaction (MFCI) leads a vapor explosion under certain circumstances. The author has performed fundamental experiment on the vapor explosion with a mass of grains of certain particle sizes which simulate the molten fuel fragments, to verify the relation between the particle size and the magnitude of pressure pulses.

The standard temperatures of water and liquid nitrogen used as cold liquid set prior to the test are 25°C and 77K respectively, and that of grains (Sic) are 600°C for the former and 25°C for the latter experiments. For both experiments, the maximum pressure pulse has the greatest value at the grain size of 0.27mmφ. This value of diameter agrees approximately with the median of the size distribution of the fragments measured in some vapor explosion experiments with a hot molten metal.

In the results of using water as cold liquid, boiling pressure traces show the oscillations of higher frequency than 100 Hz with particle sizes ranging 0.2–0.5 mm. The initial temperatures of grains and water little effect on generating such oscillations as far as it is tested in this study.     

Fuel Fragmentation and Mechanical Energy Conversion Ratio at Rapid Deposition of High Energy in LWR Fuels

The fuel fragmentation is one of the important subjects in the field of molten fuel-coolant interaction (MFCI) since it is one of basic processes of the MFCI, and it has not yet been made clear enough. Accordingly, U02 fuel fragmentation was studied in a postulated reactivity initiated accident (RIA) condition by the Nuclear Safety Research Reactor (NSRR). The distribution of the size of fuel fragments was obtained through the experiments and the mechanism of fuel fragmentation was studied. Also, the relation between the conversion ratio of the mechanical energy to the thermal and the degree of fuel fragmentation was obtained experimentally.

It was revealed that the distribution of fuel fragments was well described in the form of logarithmic Rosin-Rammler's distribution law. The fuel fragmentation was found to be explained by the Weber-type hydraulic instability model and the internal pressurization model. It was also shown that the mechanical energy conversion ratio was inversely proportional to the volume-surface mean diameter defined as the ratio of the total volume of fragments to the total surface, and furthermore that it was influenced by the coolant subcooling and the volumetric ratio of fuel to water.     

Experimental investigation on the bursting of single molten droplet in coolant

An experiment facility for observing low-temperature molten tin alloy droplet into water was es- tablished to investigate mechanisms of vapor explosion occurring in severe accidents of a fission nuclear reactor.The vapor explosion behaviors of the molten material were observed by a high-speed video cam- era and the vapor explosion pressures were recorded by a pressure transducer mounted under the water surface.The results showed that the pressure reached a peak value when the molten metal temperature was 600℃-650℃,and the coolant temperature had an obvious decreasing effect on the droplet breakups.A model for single droplet fuel/coolant interaction is proposed.It considers that in the case of Rayleigh-Taylor instability,the coolant that jets from opposite direction penetrates into the fuel and the vapor explosion occurs because of the rapid evaporation.This model explained the effect of metal droplet temperature and coolant temperature on vapor explosion.     

Transient deformation of LMFBR cores due to local failure: Experimental and theoretical investigation

This paper describes an effort to predict the mechanical core deformation caused by local failure within an LMFBR core. These activities are intended to cover all the potential core damage possibilities currently under discussion and analysis. In particular it is shown that the reactor can be scrammed in time under pessimistic-realistic pressure transients and that the damage does not exceed tolerable limits.A special gas generator technique to simulate a fuel coolant explosion has been developed at AWRE Foulness. This has been used to perform the explosion tests needed to demonstrate the safety of the SNR 300 core. A molten fuel—coolant interaction (MFCI) experimental facility, and a drop tower to carry out sub-assembly crushing tests are described. Theoretical studies are presented which use mass-spring-dashpot, lumped parameter-hinge or micro-rigid-lumped-mass models. They simulate the crushing and bending of a single sub-assembly interacting with the coolant as well as the behaviour of a multirow “spoke” model.For the core analysis only preliminary computational results are presently available which can be compared with the full scale tests in which the fluid pressure did not exceed a “threshold” of about 100 bar. Parameter studies show the influence of pulse shape, material properties as well as the time integrator.Some of the unanswered question concern the dydrodynamic feedback of the deformations on the pressure distribution in space and time. Also the behaviour of the highly irradiation-embrittled cores is poorly understood today. Finally, an enhanced energy release package to describe the MFCI must still be added to the reactivity calculation module of a future fast reactor dynamics code.     

A description of a fuel-coolant thermal interaction model with application in the interpretation of experimental results

A coherent thermo- and hydrodynamic model to explain molten fuel-coolant interactions in the CORECT II experimental facility is described. The effects of heat losses to the surroundings, mass entrainment, two-phase frictional losses and non-condensable gases are taken into account for the coolant modelization, while a solidification-controlled fragmentation model is proposed for the fuel behavior. Interpretation of two significant experiments in the CORECT II facility permit verification of the consistency of the formulation for the behavior of the fuel-coolant couple. The interpretations allow a generalized interaction scenario in this facility to be proposed, illustrate the importance of understanding the mechanisms underlying the prefragmentation and mixing phase preceding the thermal interaction, and confirms the hypothesis that the process of solidification in the fuel is a major factor governing fuel fragmentation.     

Experimental and computational study on jet final depth under coolant jet-melt interaction

Pouring coolant into molten material provides an efficient method for cooling molten core debris in light water reactor. This coolant jet-melt interaction mode needs to be studied for proposed application and safety concern. The jet breakup pattern and its final depth are crucial factors for efficient cooling. In the present study, the hydraulic penetration behavior of coolant jet is investigated using experimental and numerical approaches. A series of visual experiments are conducted using low-density gasoline as coolant jet and high-density water as molten fuel. The images of jet behaviors and the data of gasoline jet penetration depth are obtained and analyzed. Based on FLUENT15.0 a 3D axisymmetric model is built and Volume of Fluid (VOF) method is used. The hydraulic penetration behaviors of jet and final penetration depth are correctly simulated and analyzed. The fluctuating phenomenon of penetration depth and the effects of various parameters are discussed. Jet velocity and density ratio are key factors to final penetration depth. The conclusions are helpful to understand the parameter influence and the fluctuation mechanism of final penetration depth and substantiate the understanding of the coolant jet hydraulic penetration behavior during coolant jet-melt interaction.     

Energetics of simulated HCDA bubble expansions: Some potential attenuation mechanisms

Experiments conducted to increase our understanding of the dynamics and thermodynamics of expanding bubbles similar to the core disruptive accident (CDA) bubble in liquid metal fast breeder reactors (LMFBR) are described. The experiments were conducted in a transparent model of a typical demonstration-size loop-type LMFBR in which water at room temperature simulated the sodium coolant. Nitrogen gas (1450 psia) and flashing water (1160 psia) qualitatively simulated sodium vapor and molten fuel expansions. Three physical mechanisms that may result in attenuation of the work potential of a hypothetical CDA (HCDA) were revealed by the experiments: (1) the pressure gradient existing between the lower core and the bubble within the pool, (2) the hydrodynamic effects of vessel internal structures, and (3) the nonequilibrium flashing process occurring in the lower core. These three mechanisms combine to result in a coolant axial slug kinetic energy that is only 14% of the work potential of the ideal quasi-static nitrogen expansion and only 5% of the work potential of the ideal quasi-static flashing water expansion.     

熔融液滴在冷却剂中的运动特性实验

设计、建立了研究高温熔融液滴与冷却剂相互作用的可视化实验装置,通过高速摄影记录熔融液滴的下落过程,获得了下落小球运动过程曲线,重点考察了液滴温度和冷却剂温度对液滴-冷却剂界面作用过程的影响.结果表明,熔融液滴穿过气-水界面后,将首先经历一个速度骤降-回升过程,之后液滴作减速运动下落;当冷却水温度一定时,高温熔融液滴温度越高,熔融液滴入水后下落速度越快;当熔融液滴温度一定时,冷却水温度越高,熔融液滴入水后下落速度越快.     

Outline of a new thermodynamic model of energetic fuel-coolant interactions

Contact between overheated nuclear fuel and cooler volatile liquid coolant can, in certain circumstances, result in the explosive generation of vapour. Such events, known as energetic fuel-coolant interactions (FCIs), are believed to have been responsible for the ruptures of fuel channels that led to the destruction of the core of Chernobyl Unit 4. This paper outlines a new thermodynamic model of energetic FCIs, in which the response of fluids around the interaction zone is treated explicitly. By assuming that these fluids are compressed reversibly and adiabatically, a qualified lower limit to the efficiency of conversion of thermal energy to mechanical work is obtained. Furthermore, the efficiencies predicted by this new model are found to be in good agreement with those determined from the scale-urania-water (SUW) experiments at the Atomic Energy Establishment, Winfrith.     

A thermal stress mechanism for the fragmentation of molten UO2 upon contact with sodium coolant

An important aspect of fuel-coolant interaction problems relative to various hypothetical LMFBR accidents is the fragmentation of molten oxide fuel on contact with sodium coolant. In order to properly analyze the kinetics of such an event, an understanding of the breakup process and an estimate of the size and dispersion of such fragmented fuel must be known. A thermal stress initiated mechanism for fragmentation is presented using elastic stress theory for the cases of both temperature-dependent and independent mechanical properties. Included is a study of the effect of the choice of surface heat transfer boundary condition and the compressibility of the unsolidified inner core. Results of parametric calculations indicate that the thermal stresses induced in the thin outer shell and the pressurization of the inner molten core are potentially responsible for the fragmentation. For UO₂ in Na the calculated stresses are extremely high, while for aluminum in water they are much smaller and a strong function of the surface heat transfer boundary condition. Qualitatively, these results compare favorably with small scale dropping experiments, that is, molten UO₂ quenched in Na undergoes fragmentation while aluminum in water usually results in little breakup. The experimentally observed increase in breakup with decreasing coolant temperature is also in qualitative agreement with the thermal stress-induced mode of fragmentation.     

A subchannel analysis code MATRA-LMR for wire wrapped LMR subassembly

In sodium cooled liquid metal reactors design limits are imposed on the maximum temperatures of the cladding and fuel pins. Thus an accurate prediction of the core coolant/fuel temperature distribution is essential to LMR core thermal hydraulic design. The detailed subchannel thermal hydraulic analysis code MATRA-LMR is being developed for LMFBR core design and analysis based on COBRA-IV-I and MATRA. The major modifications and improvements implemented in MATRA-LMR are as follows: sodium property calculation subprogram, sodium coolant heat transfer correlations, and most recent pressure drop correlations. To assess the development status of this code, benchmark calculations were performed with the ORNL 19 pin tests and EBR-II seven-assembly SLTHEN calculation results. The calculation results of MATRA-LMR were compared to the measurements and to the SABRE4 and SLTHEN code calculation results, respectively. Finally, the major technical results of the conceptual design for the KALIMER U-10%Zr binary alloy fueled core have been compared with the calculations of the MATRA-LMR, SABRE4 and SLTHEN codes.     

Safety analysis of an accelerator-driven test facility

One of the milestones in the roadmap of accelerator-driven transmutation of waste (ATW) of the U.S. Department of Energy is the design and construction of an accelerator-driven test facility (ADTF) with a thermal power of 100 MW. Analysis of the dynamic behavior of the ADTF has been carried out in the frame of a bilateral collaboration between the Forschungszentrum Karlsruhe and the Argonne National Laboratory (ANL). In the present study five different system configurations with various types of fuel and different types of coolant have been taken into consideration.In the systems with sodium as coolant, the transient behavior under the unprotected loss-of-flow scenario shows the most serious safety concern. As long as the external source is switched on, loss-of-flow will lead to an overheating of coolant, cladding and fuel. Boiling of coolant, cladding failure and molten fuel injection take place just in several seconds after the coast-down of the pump. Safety measures have to be designed for switching off the proton beam.In the system with liquid lead–bismuth eutectic (LBE) as coolant, the buoyancy effect is much stronger. Due to its high boiling point, coolant boiling and, subsequently, flow oscillation in fuel assemblies can be avoided. By a proper design of the heat removal system, the buoyancy-driven convection would provide a sufficiently high cooling capability of the reactor core, to keep the integrity of the fuel pins.     

Experimental study on fuel-discharge behaviour through in-core coolant channels

In core-disruptive accidents of sodium-cooled fast reactors, fuel discharge from the core region reduces the possibility of severe re-criticality events. In-core coolant channels with large hydraulic diameters, such as the control-rod guide tube and a concept of the Fuel Assembly with Inner Duct Structure have a potential to provide effective fuel-discharge paths if effects of sodium in these paths on molten fuel discharge are limited. Two series of experiments were conducted to investigate fuel-discharge behaviour through the sodium-filled channels. In the first series of experiments, an alloy with low melting temperature was ejected into a water channel to clarify dominant phenomena for melt discharge through the coolant-filled channel and to develop methodologies for evaluating the effects of coolant on melt discharge. In the second series of experiments, a molten alumina was discharged through the sodium-filled channel in order to verify the applicability of the knowledge and evaluation methodologies obtained in the first series of experiments to the sodium-filled channel. These series of experiments showed that the discharge path can be entirely voided by the vaporisation of a part of the coolant at the initial melt discharge phase that this is followed by coolant vapour expansion and that melt penetrates significantly into the voided channel. Preliminary extrapolation of the present results to the in-core coolant channel suggests that the effects of the sodium on fuel discharge are limited and, therefore, in-core coolant channels will provide effective fuel-discharge paths for reducing neutronic activity.     

Effect of Volumetric Ratio and Injection Pressure on Water-Liquid Nitrogen Interaction

To simulate the fuel-coolant interaction under the easily controllable condition, the water and the liquid nitrogen were used respectively as the molten fuel and the coolant in this study. To initiate the interaction, the water was injected into the cylindrical chamber to come into contact with the liquid nitrogen. The experiments were conducted with two key parameters, the water/liquid nitrogen volumetric ratio and the water injection pressure, to study their effects on the interaction. The experiments were conducted at the different water injection pressures with various initial volumetric ratios for water and liquid nitrogen. The pressure spikes and the ice debris observed from the experiments confirmed the existence of the strong interaction between the water and the liquid nitrogen. The occurrence of the pressure spikes and the times of their inceptions at the different conditions were used to create a diagram to indicate the preferable conditions for the strong interaction. The propagation speed of an observed pressure spike was estimated to be comparable to the theoretical sound speed of the liquid nitrogen/nitrogen vapor mixture. This result suggested the possibility of the vapor explosion interaction between the water and the liquid nitrogen similar to that observed in the fuel-coolant interaction.     

高温小球在冷却剂中运动阻力特性的研究

通过实验．研究了高温小球在冷却剂中的运动阻力特性,揭示出高温难挥发的熔融物与低温易挥发的冷却剂相互作用的粗混合阶段具有特殊结构高温熔融金属液滴在冷却剂中的运动规律。研究结果对于开发多成分多相分析程序和研究核电站严重事故中燃料与冷却剂的相互作用具有重要意义一