Transformation and fragmentation behavior of molten metal drop in sodium pool

In order to clarify the fragmentation mechanism of a metallic alloy (U–Pu–Zr) fuel on liquid phase formed by metallurgical reactions (liquefaction temperature = 650 °C), which is important in evaluating the sequence of core disruptive accidents for metallic fuel fast reactors, a series of experiments was carried out using molten aluminum (melting point = 660 °C) and sodium mainly under the condition that the boiling of sodium does not occur. When the instantaneous contact interface temperature (T_i) between molten aluminum drop and sodium is lower than the boiling point of sodium (T_c,bp), the molten aluminum drop can be fragmented and the mass median diameter (D_m) of aluminum fragments becomes small with increasing T_i. When T_i is roughly equivalent to or higher than T_c,bp, the fragmentation of aluminum drop is promoted by thermal interaction caused by the boiling of sodium on the surface of the drop. Furthermore, even under the condition that the boiling of sodium does not occur and the solid crust is formed on the surface of the drop, it is confirmed from an analytical evaluation that the thermal fragmentation of molten aluminum drop with solid crust has a potential to be caused by the transient pressurization within the melt confined by the crust. These results indicate the possibility that the metallic alloy fuel on liquid phase formed by the metallurgical reactions can be fragmented without occurring the boiling of sodium on the surface of the melt.     

A molten metal jet impingement on a flat spreading surface

ABSTRACT

In the event of a severe accident, past experiences such as Three Mile Island and Fukushima Daichi have shown that the reactor core of a light-water nuclear reactor, if not properly safeguarded, could go through a meltdown. This will be followed by the formation of a corium, a mix of molten fuel elements, and liquid metals from the Reactor Pressure Vessel (RPV). In the worst-case scenario, a melt through from the RPV can occur and lead to the spreading of the corium, in the form of a molten element’s jet impinging on a flat concrete structure of the Primary Containment Vessel (PCV). To enhance the decommissioning and the safety procedure, scope of the present article is to deepen the understanding of the phenomena involved in the mentioned scenario, mainly jet-instability and molten material spreading. In the present study, experiments were carried out, by using corium simulant materials such as Copper and Tin, to investigate the link between the instability of the gravity-driven molten metal jet and the impinging followed by its spreading over a flat area.     

Study on thermal-hydraulic behavior during molten material and coolant interaction

In a core disruptive accident (CDA) of a Fast Breeder Reactor, the post accident heat removal (PAHR) is crucial for the accident mitigation. The molten core material should be solidified in the sodium coolant in the reactor vessel. The material, being fragmented while solidification and forming debris bed, will be cooled in the coolant.

In the experiment, molten material jet is injected into water to experimentally obtain the visualized information of the fragmentation and boiling phenomena during PAHR in CDA. The experiment shows that the break up of the molten material into fine fragments is observed at the front, side and middle part of the jet during very short time interval. The distributed particle behavior of the molten material jet is observed with high-speed video camera. And the visual data is analyzed with Particle Imaging Velocimetry (PIV).

The experimental results are compared with the existing theories. Consequently, the marginal wavelength on the surface of a water jet is close to the value estimated based on the Rayleigh–Taylor instability. Moreover, the fragmented droplet diameter obtained from the interaction of molten material and water is close to the value estimated based on the Kelvin–Helmholtz instability.     

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.     

On the explosivity of a molten drop submitted to a small pressure perturbation

Under some circumstances that we aim to determine, a hot molten fuel drop flowing into a volatile liquid coolant and submitted to a small pressure wave, can be destabilized and explode in a few milliseconds. We propose a new approach to address this problem: in contrast with previous studies, we do not try to model the complete phenomenon but concentrate on its initiation. This way, we can differentiate favourable and non-favourable conditions with applications to PWR/BWR safety. We do the hypothesis that the occurrence of contacts between the two fluids is the criterion of explosion and phenomena occurring up to the contacting event are modelled, including the vapour film oscillations and the amplification of Rayleigh-Taylor instabilities at its interface. The latter feature receives a particular attention with a transient modelling adapted for variable acceleration cases. The fragmentation process itself is not studied in details but we give arguments supporting the fact that contacts between both liquids should induce a strong destabilization of the drops and initiate fragmentation. In this way, we can characterize the explosivity, i.e. the ability for the drop to explode, as a function of the various physical properties (e.g. pressure, temperatures). The model so deduced is qualified by comparison with the explosivity maps provided by Nelson and Duda [Nelson, L.S., Duda, P.M., 1985. Steam explosion experiments with single drops of iron oxide: Part II: parametric studies, NUREG CR-2718, April 1985]. Results obtained with this model confirm the experimental trends regarding the role of ambient pressure and liquid temperature. The influence of other parameters as the drop and the trigger characteristics are also investigated. We conclude this paper with some consideration on the implications for nuclear safety.     

Experimental investigation of solid–liquid mixtures freezing behavior in flow channels

The penetration and freezing of hot-core material mixtures through flow channels during core disruptive accidents (CDAs) within a sodium-cooled fast reactor is one of the major concerns confronting safety designers of the next-generation reactors. The main objective of this study is to investigate those fundamental characteristics of penetration and solidification involved in channeling molten metal and solid particle mixtures over cold structures. In this study, a low-melting-point alloy (viz., Bi–Sn–In alloy) and mixtures with solid particles (of copper and bronze) were used as a simulant melt, while L-shape metal (of stainless steel and brass) and stainless steel fuel pin bundle were used as cooling structures. Two series of basic experiments were performed to study the effect solid particles have on penetration and cooling behavior under various thermal conditions of melt by varying solid particle volume fraction, structure temperature and structural geometry. Melt flows and distributions were recorded using a digital video camera and subsequently analyzed. The melt penetration length into the flow channel and the proportion of melt adhesion on structural surfaces were measured. Results indicate that penetration length becomes shorter for molten-metal/solid particle mixtures (mixed melts) compared with pure molten metal (pure melt) as well as decreases with increasing solid particles volume fraction of the melt. The present study will contribute to a better understanding of the basic thermal-hydraulic phenomena of melt freezing in the presence of solid particles and to provide an experimental database for validation and improvement of the models of fast reactor safety analysis codes.     

Corrosion behavior of Ni-based structural materials for electrolytic reduction in lithium molten salt

In this study, the corrosion behavior of new Ni-based structural materials was studied for electrolytic reduction after exposure to LiCl-Li₂O molten salt at 650 °C for 24-216 h under an oxidizing atmosphere. The new alloys with Ni, Cr, Al, Si, and Nb as the major components were melted at 1700 °C under an inert atmosphere. The melt was poured into a preheated metallic mold to prepare an as-cast alloy. The corrosion products and fine structures of the corroded specimens were characterized by scanning electron microscope (SEM), Energy Dispersive X-ray Spectroscope (EDS), and X-ray diffraction (XRD). The corrosion products of as cast and heat treated low Si/high Ti alloys were Cr₂O₃, NiCr₂O₄, Ni, NiO, and (Al,Nb,Ti)O₂; those of as cast and heat treated high Si/low Ti alloys were Cr₂O₃, NiCr₂O₄, Ni, and NiO. The corrosion layers of as cast and heat treated low Si/high Ti alloys were continuous and dense. However, those of as cast and heat treated high Si/low Ti alloys were discontinuous and cracked. Heat treated low Si/high Ti alloy showed the highest corrosion resistance among the examined alloys. The superior corrosion resistance of the heat treated low Si/high Ti alloy was attributed to the addition of an appropriate amount of Si, and the metallurgical evaluations were performed systematically.     

纯镍渗硼在熔融LiCl-Li2O中的腐蚀行为

采用固体粉末包装法在950℃渗硼5 h,在纯镍表面制备了厚度约50μm致密连续的渗硼涂层.采用浸没法研究了纯镍及纯镍渗硼涂层在750℃空气及熔融LiCl-l0Li2O(Li2O质量分数为l0%)中的腐蚀行为.渗硼涂层试样的腐蚀失重为8.4 mg·cm-2,只有原始纯镍试样的1/4.渗硼涂层明显改善了纯镍在熔融LiCl-Li2O中的耐腐蚀性能,这归因于渗硼涂层中的硼元素在熔融LiCl-Li2O中的优先腐蚀抑制了镍的加速腐蚀.     

Numerical investigation of ductile crack growth behavior in a dissimilar metal welded joint

In this paper, the finite element method (FEM) based on GTN model is used to investigate the ductile crack growth behavior in single edge-notched bend (SENB) specimens of a dissimilar metal welded joint (DMWJ) composed of four materials in the primary systems of nuclear power plants. The J-Δa resistance curves, crack growth paths and local stress-strain distributions in front of crack tips are calculated for eight initial cracks with different locations in the DMWJ and four cracks in the four homogenous materials. The results show that the initial cracks with different locations in the DMWJ have different crack growth resistances and growth paths. When the initial crack lies in the centers of the weld Alloy182 and buttering Alloy82, the crack-tip plastic and damage zones are symmetrical, and the crack grow path is nearly straight along the initial crack plane. But for the interface cracks between materials and near interface cracks, the crack-tip plastic and damage zones are asymmetric, and the crack growth path has significant deviation phenomenon. The crack growth tends to deviate into the material whose yield stress is lower between the two materials on both sides of the interface. The different initial crack locations and mismatches in yield stress and work hardening between different materials in the DMWJ affect the local stress triaxiality and plastic strain distributions in front of crack tips, and lead to different ductile crack growth resistances and growth paths. For the accurate integrity assessment for the DMWJ, the fracture toughness data and resistance curves for the initial cracks with different locations in the DMWJ need to be obtained.     

Simulation of the physical effects of the interaction of molten fuel with the coolant

Conclusions The method proposed makes it possible to obtain computational estimates of the intensity of a steam explosion inside a reactor vessel and in the space below the reactor inside the melt trap. The computational investigations of the intensity of a steam explosion inside a VVéR vessel in the most likely scenario of a serious accident with efflux of melt into the bottom pressurized chamber show that under certain conditions a high pressure capable of destroying separate structural elements can develop. The mass of the interacting melt, the initial temperature, the fragmentation time, and the final size of the fragments, as well as the type of contact realized, have the greatest effect on the intensity of the steam explosion. Local steam explosions in pipes of the melt trap have a relatively low intensity and cannot have a large effect on the construction in the space below the reactor and on the containment envelope. Deceased. State Science Center of the Russian Federation — Physics and Power Engineering Institute. Translated from Atomnaya énergiya, Vol. 80, No. 1, pp. 3–10, January, 1996.     

Electrochemical behavior and electronic absorption spectra of uranium trivalent ions in molten LiCl-CsCl mixtures

The redox reactions and coordination circumstances of uranium trivalent ions in molten LiCl-CsCl mixtures were investigated by cyclic voltammetry and spectrophotometry. The formal redox potential, E°′(U³⁺|U), in LiCl-CsCl mixtures with the CsCl mole fraction of 0.2 was more positive than that in LiCl melt. The CsCl system showed the most negative E°′(U³⁺|U). The electronic absorption spectra of U³⁺ in LiCl-CsCl mixtures showed that the intensities of absorption peaks decreased with the increase of CsCl mole fraction. The oscillator strength of the hypersensitive f-f transition, f, decreased with the increase of CsCl mole fraction.     

Separation of Nd metal by using disproportionation reaction of Nd(II) in molten chlorides

In order to separate neodymium (Nd) from lanthanides in chloride melts, the electrochemical characteristics Nd ions in molten LiCl-CaCl₂ eutectic were studied. The formal redox potentials of the Nd³⁺|Nd²⁺ and Nd²⁺|Nd couples in molten LiCl-CaCl₂ eutectic at 823 K were determined to be −2.745 ± 0.005 and −3.081 ± 0.005 V vs. Cl₂|Cl⁻. Under the controlled potential electrolysis by applying negative potential to form Nd²⁺, Nd²⁺ was disproportionated to Nd³⁺ and metallic Nd fog according to the reaction; 3Nd²⁺ ? 2Nd³⁺ + Nd. When a quartz glass was immersed in the melt during the electrolysis, Nd was coated on the quartz surface. The chemical composition of the recovered Nd was analyzed to be Nd metal and Nd₂O₃ by scanning electron microscopy, X-ray diffractometry, and electron probe microanalysis. The same electrolytic method was carried out under the coexistence of Nd³⁺ and lanthanum ion (La³⁺). Nd³⁺ was separated from La³⁺ and recovered to be Nd₂O₃.     

Analysis on the neutron kinetics for a molten salt reactor

The neutron kinetics of the molten salt reactor is significantly influenced by the fuel salt flow, which leads to the analysis methods for the conventional reactors using solid fuels not being applicable for the molten salt reactors. In this study, a neutron kinetic model considering the fuel salt flow is established based on the neutron diffusion theory, which consists of two-group neutron diffusion equations for the fast and thermal neutron fluxes and six-group balance equations for delayed neutron precursors. The temperature feedback on the neutron kinetics is considered by introducing a heat transfer model in the core, in which the group constants which are dependent on the temperature are calculated by the code DRAGON. The mathematical equations are discretized and numerically calculated by developing a code, in which the fully implicit scheme is adopted for the time-dependent terms, and the power law scheme is for the convection–diffusion terms. The neutron kinetics is conducted during three transient conditions including the rods drop transient, the pump coastdown transient and the inlet temperature drop transient. The relative power changes and the distributions of the temperature, neutron fluxes and delayed neutron precursors under these three different transient conditions are obtained in the study. The results provide some valuable information for the research and design of this new generation reactor.     

Thermo-mechano-chemical stability of ceramic materials during the electrowinning process using liquid metal electrodes in molten salts

Pyroprocessing, which results in proliferation resistance, shows promise as an alternative to wet processing in the recycling of transuranics. However, the ceramic crucible used in the electrowinning process poses an issue during pyroprocessing. The crucible is chemically unstable and prone to thermal fatigue. In this study, the thermodynamic simulation software HSC (enthalpy, entropy and heat capacity) Chemistry was employed to evaluate the chemical stabilities of different ceramic crucibles containing liquid cadmium as well as liquid bismuth cathodes, which also contained rare earth elements and lithium. The chemical stabilities were experimentally validated by measuring the contact angles between the liquid cathode (LC) materials and four ceramic materials (Al₂O₃, MgO, Y₂O₃, and BeO) in situ. The infiltration depths of the liquid bismuth cathode elements were measured using X-ray photoelectron spectroscopy. To determine the Weibull distributions of the investigated ceramics, thermal fatigue tests were performed using plates of the ceramics.     

Simulation of high-temperature swelling of uranium dioxide and deformation behavior of a fuel element

Models and computer codes, developed based on them, for simulating the swelling of uranium dioxide (BARS) and the stress-deformation state of a fuel element (SDS) under high-temperature irradiation are presented. It is shown that when developing a design for high-temperature fuel elements and validating their serviceability the quantitative indicator required for the swelling of uranium dioxide in the range ≥1400°C is the change in the external dimensions of the fuel caused by constant formation and growth of bubbles containing gaseous fission products during irradiation. The results of computational investigations using the models indicated are examined. These results eliminate the inconsistency of the data on the effect of the main operating parameters — the temperature and burnup — on the radiation characteristics and service life behavior of a fuel element. It is shown that the central channel in the fuel kernel and strengthening of the cladding improve the dimensional stability fuel elements. __________ Translated from Atomnaya énergiya, Vol. 103, No. 3, pp. 172–179, September, 2007.     

Simulation of stress corrosion cracking behavior in a tube-shaped specimen of nickel-based alloy 600

Stress corrosion cracking (SCC) simulation code has been developed for the evaluation of SCC behavior in specimens in the shape of field components. The code utilizes numerical calculation of stress/strain states at a crack tip using finite element methods and a formula describing the crack tip reaction kinetics containing unknown environmental parameters. The applicability of this simulation code was investigated by applying the code to the evaluation of SCC behavior in a mock-up of a bottom mounted instrumentation tube for a pressurized water reactor subjected to complex stress/strain states. The results indicate that crack growth rate in a component suffering from certain environments can be estimated using the developed SCC simulation code with pre-determined unknown parameters, using the experimental crack growth rate data measured on other specimens in the same environment.     

Fragmentation of a molten corium jet falling into water

During a severe accident of a pressurized water nuclear reactor, a large mass of corium could pour into the vessel bottom as a compact jet. When the corium mass reaches the water at the bottom of the vessel, an intense fragmentation may occur. This could lead to a significant mixing of corium and water, likely to cause a steam explosion which could damage the structures. An analytical study has been established in order to quantify the corium jet fragmentation. This study consists mainly in modeling the vapor flow surrounding the jet as well as the instability which occurs at its interface. In comparison with previous studies, this model pays particular attention to the jet-produced particles which interact with the vapor flow. A complete model has been set up in order to calculate the jet breakup length and the generated particles’ diameter under each specific situation characterized by initial conditions. This model mainly relies upon results from boundary layer theory and linear instability calculations. The full model’s results are compared to existing experiences in this field and a final correlation of the results is established. A good agreement is obtained on the jet breakup length, however the predicted particle diameter tends to be too large. This last result could be explained by a secondary breakup of the particles in water and by a large uncertainty in the vapor flow.